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First plasma operation in ITER will start after completing the assembly of the tokamak vessel and the installation of the main sub-systems, but prior to the installation of the blanket modules and the divertor cassettes. Utilization of temporary limiters and divertor replacement structures will provide a poloidal and toroidal reference side position to the plasma edge to protect the vacuum vessel and other components already installed during operations. An additional set of mirrors is required to reflect the power injected from the upper launcher towards the plasma resonance for EC-assisted breakdown of the plasma up to a beam dump needed to trap and absorb the power of the beams after crossing the plasma in order to reduce the stray radiation escaping back into the vacuum chamber down to less than 10% of the total power. The quasi-optical system has been designed, with shape and size of the mirrors compliant with the requirements provided by ITER for their installation, realization and plasma performances, resulting in two standard focussing mirrors and one grating mirror. The beam dump consists of a box with five metal plates, the first providing a spreading of the high incident power and the others coated with absorbing material with thickness distribution studied to gradually reduce the power during the multiple reflections inside the box, avoiding damages to the coating itself. This work focuses on the validation of the quasi-optical design of the mirrors and the assessment of the dump performances, based on a multi-bounces model developed ad-hoc for this purpose. The study includes a tolerance analysis for the beam dump to include the effect of uncertainties in the thickness of the absorbing coating and misalignments of the mirrors, to verify the performances of the dump also when operating in different conditions with respect to the nominal ones.

The ITER first plasma operations will occur with no full blanket and divertor components installed. Machine protection is required and First Plasma Protection Components (FPPC) have been designed to shelter the vacuum vessel and other components from the plasma itself and from high power Electron Cyclotron Resonance Heating (ECRH) foreseen for first plasma breakdown. ECRH protection components will be installed to protect in-vessel structures from direct and stray radiation as EC beams will be used for plasma breakdown. This paper is focused on the design of these components including dedicated mirrors to shape and redirect the beams to the EC resonance location in the magnetic field null region and then into a beam dump located in an equatorial port, where exceeding EC radiation will be trapped and dumped. Two mirrors and one grating mirror have been designed to provide the required shaping and directions for the beams coming from the upper launcher towards EC resonance and dump. The beam dump consists of five large plates affixed in an equatorial port. Guidelines that drove the design of the quasi-optical system, characteristics of the mirrors, resulting launched beams and concept developed for the beam dump will be here described.

High-power microwaves generated by gyrotrons will be extensively used in ITER for a variety of purposes such as assisting plasma breakdown, plasma heating, current drive, tearing mode suppression and as a probing beam for the Collective Thomson Scattering diagnostic. In a number of these schemes absorption of the microwaves by the plasma will not be full and in some cases there could be no absorption at all. This may result in a directed beam with a high microwave power flux or – depending on location and plasma conditions – an approximately isotropic microwave power field. The contribution of electron cyclotron emission to these power densities is briefly discussed. Exposure to in-vessel components leads to absorption by metals and ceramics. In this paper microwave power densities are estimated and, following a brief review of absorption, thermal loads on in-vessel components are assessed. The paper is concluded by a discussion of the current approach to control such loads.

The electron cyclotron (EC) heating and current drive (H&CD) system developed for the ITER is made of 12 sets of high-voltage power supplies feeding 24 gyrotrons connected through 24 transmission lines (TL), to five launchers, four located in upper ports and one at the equatorial level. Nearly all procurements are in-kind, following general ITER philosophy, and will come from Europe, India, Japan, Russia and the USA. The full system is designed to couple to the plasma 20 MW among the 24 MW generated power, at the frequency of 170 GHz, for various physics applications such as plasma start-up, central H&CD and magnetohydrodynamic (MHD) activity control. The design takes present day technology and extends toward high-power continuous operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The development of the EC system is facing significant challenges, which includes not only an advanced microwave system but also compliance with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since the conceptual design of the EC system was established in 2007, the EC system has progressed to a preliminary design stage in 2012 and is now moving forward toward a final design.

An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion. (C) 2014 Elsevier B.V. All rights reserved.

A 24 MW Electron Cyclotron (EC) system operating at 170 GHz and 3600 s pulse length is to be installed on ITER. The EC plant shall deliver 20 MW of this power to the plasma for Heating and Current Drive (H&CD) applications. The EC system is designed for plasma initiation, central heating, current drive, current profile tailoring, and Magneto-hydrodynamic control (in particular, sawteeth and Neo-classical Tearing Mode) in the flat-top phase of the plasma. A preliminary design review was performed in 2012, which identified a need for extended application of the EC system to the plasma ramp-up, flattop, and ramp down phases of ITER plasma pulse. The various functionalities are prioritized based on those applications, which can be uniquely addressed with the EC system in contrast to other H&CD systems. An initial attempt has been developed at prioritizing the allocated H&CD applications for the three scenarios envisioned: ELMy H-mode (15 MA), Hybrid (similar to 12 MA), and Advanced (similar to 9 MA) scenarios. This leads to the finalization of the design requirements for the EC sub-systems.

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A collaborative effort between European research Associations and Thales Electron Devices (TED) has been launched by the European Fusion Development Agreement (EFDA) in 2003, aiming at the development of an industrial 170 GHz/2 MW/CW coaxial cavity gyrotron. The first prototype is expected to reach 2 MW/1s and is presently being tested in Lausanne at a dedicated test facility. The test facility has been designed to be flexible: allowing the possible commissioning of tubes with different characteristics, as well the tests of a version of the ITER upper launcher antenna at full performances. The test facility has been commissioned during the first test phases on the gyrotron which include: high voltage stand-off, coaxial insert alignment and cathode conditioning without depressed collector power supply at a slightly reduced electron beam power of 85 kV/78 A/2 ms. The short pulse (

In ITER, EC heating and current drive (H&CD) is foreseen not only as a principal auxiliary system for plasma heating and as assist for plasma start-up, but is considered essential in meeting the key requirement of neoclassical tearing mode (NTM) stabilisation, by localized current drive. In the reference ECH design, ITER requires a total of 20 MW/CW power at 170 GHz using gyrotrons with a unit power of 1 MW. A higher power per unit (2 MW/gyrotron) would result in a strong reduction of the cost of the whole ECRH system, and would also relax the room constraints on the launcher antenna design. In view of the capability of coaxial cavity gyrotrons demonstrated with short pulse experiments at FZK, the European Fusion Development Agreement (EFDA) has started in 2003 the development of an industrial 170 GHz 2 MW/CW coaxial cavity gyrotron, in a collaborative effort between European research associations CRPP/EPFL, FZK, TEKES and Thals Electron Devices (TED). The development plan includes three steps to reach successively 2 MW/1s, 2 MW/60s and finally 2 MW/CW operation. The procurement of the first prototype is in progress and it scheduled to be delivered during the first quarter of 2006. The experimental tests of the prototypes will be carried out at CRPP/EPFL, where an ITER relevant test facility is presently under construction and will be achieved during the second half of 2005. The test facility is designed to be flexible enough, allowing the possible commissioning of tubes with different characteristics, as well the tests of the launcher antenna at full performances.

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The EC system at 140 Ghz, 2 MW, being implemented on FTU tokamak for performing electron heating, profile control and current drive, is composed by four gyrotrons with 0.5 s pulse length capability, and four hybrid mirror/waveguide transmission lines. The launching system is capable of poloidal/toroidal beam steering through a set of tiltable in-vacuum mirrors, coupling to the O-mode at the fundamental resonance, with full control of the e.m. field polarization. Experiments have been made with two gyrotrons, by launching 0.8 MW to the plasma. In the high density regime (local n{sub e} from {approx_equal}0.8 up to 2 10{sup 20} m{sup -3}) the e-i energy transfer is significant, and ion heating is observed through the enhancement of the neutron emission in Deuterium plasmas. The highest electron heating is observed when on-axis ECRH is performed in the current ramp-up, sawtooth-free phase. In steady-state conditions, saw-tooth period is increased up to stabilization by off-axis ECRH. Localized ECCD is also performed to assist in the shaping of the current density profile.

Localized ECRH/ECCD has been performed on FTU with the specific aim of shaping the temperature/current density profile for controlling MHD activity. In particular, m=1 mode dynamics and stability are studied during the heating phase, when the steady-state, sawtoothing conditions of the ohmic plasma are modified by a strong off-axis electron heating. Depending on the position of the absorbing layer and on the density of the ohmic target plasma, sawteeth are temporarily suppressed, still in the presence of a slowly growing m=1 oscillation, or a complete quenching of any m=1 activity can be achieved. When direct current, driven by EC waves launched at +/-10 degrees from the perpendicular in the up-shifted scheme, is added to the pure beating effect, different shapes and periods of relaxation phenomena in the plasma core are observed. Although low values of I-ECCD/I-p are expected because of rather high values of Z(eff) during heating, (I-ECCD/I(p)approximate to 0.04 for P-EC =800KW, Z(eff)=5 and I-p=360kA), different local changes in the magnetic shear are thought to be responsible of these differences. Time-dependent calculations of current profiles, including transport and diffusion effects, and an analysis of the onset and stabilization of sawtooth oscillations based on the time evolution of sl(the magnetic shear value at q=1) compared with that of a suitable critical value are performed.

Most of the interest in ECRH experiments is linked to the high localization of EC waves absorption in well known portions of the plasma volume. In order to take full advantage of this capability a reliable code has been developed for beam tracing and absorption calculations. The code is particularly important for oblique (poloidal and toroidal) injection, when the absorbing layer is not simply dependent on the position of the EC resonance only. An experimental estimate of the local heating power density is given by the jump in the time derivative of the local electron pressure at the switching ON of the gyrotron power. The evolution of the temperature profile increase (from ECE polychromator) during the nearly adiabatic phase is also considered for ECRH profile reconstruction. An indirect estimate of optical thickness and of the overall absorption coefficient is given by the measure of the residual e.m. power at the tokamak walls. Beam tracing code predictions of the power deposition profile are compared with experimental estimates. The impact of the finite spatial resolution of the temperature diagnostic on profile reconstruction is also discussed.

The 140GHz, 2MW, 0.5s ECRH system on FTU tokamak integrates closed waveguide transmission lines (≈30 m) with quasi optical systems at both ends for efficient coupling from the 4 gyrotrons to the 4 waveguides and from these to the plasma through a single access port. Poloidal and toroidal control of the beam’s launching angles and polarization is performed without movable components close to the plasma. Most of the components of each generation and transmission system were designed to operate at a power level higher than 0.5 MW, and a possible up-grade to a full 1 MW, 0.5 s capability is discussed.