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The diagnostic shielding module consists of two black body sources and two remotely retractable lenses located in window 9 of the international thermonuclear experimental reactor ( ITER). The function of this device is to generate and guide blackbody radia-tion to calibrate the radial and oblique views of ITER electron cyclotron emission diagnosis. The stability of emitter temperature is very important for the proper operation of precision source. In order to solve the above problems, the electron cyclotron emission ( ECE) system was preliminarily designed to optimize the distance between the emitter and the thermostatic heat source. The temperature distri-bution of the different paths of the ECE emitter under the radiant heat transfer condition was obtained by ANSYS simulation analysis. The results show that when the distance between the constant heating source and the emitter is 4 mm, the temperature of the emitter is stable between 693 °C and 720 °C, and the maximum surface temperature difference is less than 10 °C, which meets the requirements and constraints of the calibration source in ITER. The research results have an important guiding role in thermal structure analysis and prototype development of electron cyclotron emission ( ECE) system under radiation heat transfer. [ABSTRACT FROM AUTHOR]

We have performed one-dimensional full particle simulations of perpendicular shocks and found that an electron cyclotron microinstability can develop in the foot during the self-reformation phase of low beta(i) supercritical shocks. The instability is excited by the beam of reflected ions interacting with the incoming electrons. It exhibits a rapid growth, and propagates along the shock normal towards upstream. This instability, which does not require high Mach number, has a frequency comparable to the electron cyclotron frequency and a wavelength shorter than the electron inertia length. It basically results from the coupling of electron Bernstein waves with an ion beam mode carried by the reflected ions. Dispersion properties in the foot are analysed. We discuss the effects of varying parameters, in particular as the fake ion-to-electron mass ratio used in the simulations converges to more realistic values. Comparison with other microinstabilities evidenced in recent full particle simulations of shocks is outlined. (C) 2005 COSPAR. Published by Elsevier Ltd. All rights reserved.

The integration of the heating and current drive (HCD) systems in the EU DEMO tokamak must address a number of issues, namely space constraints in the tokamak building, remote handling requirements, breeding blanket penetration, neutron and photon radiation shielding, compliance of penetrations of the primary vacuum with safety and vacuum criteria, and a large number of loading conditions, in particular heat, electromagnetic (EM), and pressure loads in normal and off-normal conditions. A number of pre-conceptual design options for the vacuum vessel (VV) port and the port-plug are under assessment and need to be verified against all requirements and related criteria. The identification of the functional (or physics) requirements of the HCD systems remains an on-going process during the pre-conceptual design phase, hence some initial assumptions had to be made as a basis for development of the design of the VV ports and the HCD port plugs. The paper will provide an overview of present margins in the functional/physics requirements and the rationale behind the assumptions made in order to facilitate development of the pre-conceptual design options. Furthermore it will introduce the initial design concepts of the electron cyclotron (EC) Launchers and the neutral beam (NB) injectors integrated in equatorial ports. The NB duct design in DEMO and related issues such as transmission and re-ionization losses will be also addressed. [ABSTRACT FROM AUTHOR]

Within the Power Plant Physics and Technology (PPPT) programme in the EUROfusion Consortium design activities are currently in progress for the development of a DEMOnstration Fusion Power Plant (DEMO). The design of the machine and the integration of in-vessel components require neutronics analyses to verify the tritium self-sufficiency, the shielding requirements, and the structural integrity of its components. In particular, the effect of penetrations in the blanket and in equatorial port plug introduced by the electron cyclotron (EC) heating system, namely due to openings for the antenna waveguides (WGs), were analyzed. In this study three-dimensional MCNP calculations were conducted for the pre-conceptual designs of the EC port plugs and shielding optimization were performed in order to ensure that the DEMO design limits are not exceeded. Two configurations of the EC heating system were performed using a DEMO Water Cooled Lithium Lead (WCLL) with integrated EC configurations model of a half of the sector (i.e. 10° model) and relevant ones repeated with a full sector model (i.e. 20° model) to test the reliability of results. Additionally, the effect of the radiation on the WGs closest to the plasma was analyzed as well as the impact on Tritium Breeding Ratio (TBR). [ABSTRACT FROM AUTHOR]

Overmoded corrugated waveguide is efficient at transmitting power over a large frequency bandwidth. This operational flexibility becomes important in multi-frequency systems. For 50-mm diameter aluminum corrugated waveguide nominally designed for the ITER 170 GHz ECH system, the theoretical ohmic loss of the HE11mode is around 0.3e-3 dB/m. In a possible dual-frequency system at ITER, the theoretical loss in the same waveguide increases to a manageable 0.8e-3 dB/m at 104 GHz. As examples of other wideband components, General Atomics has built several pairs of 63.5-mm diameter waveguide polarizers for the TCV tokamak's ECH transmission line that are designed to operate for 1 MW at frequencies ranging from 82.6 to 118 GHz. In addition, polarizers have been designed for ITER's 50-mm diameter transmission line. A computer code that calculates both the required mirror rotation angles and the ohmic losses predicts that these polarizers will function effectively at both 170 GHz and 104 GHz. A new class of wideband waveguide switches with rotary actuators have been supplied for the 82.6–126 GHz transmission lines at TCV, further emphasizing the broadband capabilities of corrugated waveguide components. [ABSTRACT FROM AUTHOR]

• Description of the conceptual design of the ECRH launching system for DTT. • Design changes to comply with the updated reference plasma scenario. • New steering angle to limit the interaction between first wall and launched beams. • Introduction and preliminary cooling design of waveguide sections to reduce vacuum conductance between the DTT and transmission line environments. • Design of a quasi-optical ex-vessel optics unit for the matching between the transmission line and the waveguide entrance. One of the main goals of the Divertor Tokamak Test (DTT) facility is to reach a ratio of power crossing the separatrix over the major radius of about 15 MW m −1, as the one expected in DEMO. For this purpose, up to 45 MW of external additional heating power shall be coupled to the plasma, provided by Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating and Neutral Beam Injection. The foreseen total ECRH power installed at full power shall be 32 MW, generated using 1 MW/170 GHz gyrotrons, for 100 s. The present ECRH system foresees two launching antennas per DTT sector, based on the front-steering concept. The equatorial antenna is dedicated to plasma heating and current drive and the upper antenna to the stabilization of MHD instabilities. This paper focuses on the latest design concept for these two antennas and on the definition of the ex-vessel matching optics unit of the last section of the evacuated Transmission Line (TL). The design has been updated to be compatible with the insertion of CVD diamond windows, to separate the vacuum environment of DTT from the one of the TL. This choice requires adding corrugated waveguide sections between the last mirror of the TL and the first mirror inside the port, requiring some adaptation of the in-vessel optics and of the supporting structure. The possibility to modify the steering range for the launching mirror has been also investigated to be compatible with the new design of the first wall and for the upper antenna, to reach the q = 2 surface in the new plasma scenario. [ABSTRACT FROM AUTHOR]

The JT-60SA tokamak is scheduled to start operations in 2019 to support the ITER experimental programme and to provide key information for the design of DEMO scenarios. The device will count on ECRH and NBI as auxiliary heating. Potentially dangerous situations that could harm the tokamak structures during operations characterized by low EC absorption cannot be a priori excluded. An estimation of the wall load under direct beam exposure and during plasma operations will be given, using the EC antenna main characteristics and launching geometry. A system capable to prevent the risk of damage to machine components for JT-60SA is presented and examples of potential benefits of such a system will be given, based on recent experience gained on the TCV tokamak during EC wall cleaning experiments performed in preparation of JT-60SA operations. [ABSTRACT FROM AUTHOR]

The physics requirements of the heating and current (H&CD) systems in a Demonstration Fusion Power Plant (DEMO) are often beyond the actual level of design maturity and technology readiness required. The recent EU fusion roadmap advocates a pragmatic approach and favours, for the initial design integration studies, systems to be as much as possible, extrapolated from the ITER experience. To reach the goal of demonstrating the production of electricity in DEMO with a closed fuel cycle by 2050, one must ensure reliability, availability, maintainability, inspectability (RAMI) as well as performance, efficiency and optimized design for the H&CD systems. In the recent Power Plant Physics & Technology (PPP&T) Work Programme, a number of H&CD studies were performed. The four H&CD systems Neutral Beam (NB) Injection, Electron Cyclotron (EC), Ion Cyclotron (IC) and Lower Hybrid (LH) were considered. First, a physics optimization study was made assuming all technologies are available and identifying which parameters are needed to optimize the performance for given plasma parameters. Separately, the (i) technological maturity was considered (e.g. 240 GHz gyrotrons for EC) and (ii) technologies were adapted (e.g. multi-stage depressed collector for EC) or (iii) novel solutions (e.g. photo-neutralization for NB or new antennae concepts for IC) were studied to overcome the limitations of the present H&CD systems with respect to DEMO requirements. Further constraints imposed by remote maintenance or breeding blanket interactions were considered. [ABSTRACT FROM AUTHOR]

The design of a high voltage amplifier was being developed that uses series-connected transistors to control the output [Tooker and Huynh, 2013]. Although having many potential applications, this amplifier is designed to meet at a minimum the requirements of a body power supply for depressed collector gyrotrons that are used in electron cyclotron systems on the DIII-D tokamak and other devices worldwide. Therefore, the output voltage needs to be adjustable up to 35 kV DC and be capable of square-wave modulation at frequencies up to 5 kHz. Continuous operating body currents for a gyrotron are on the order of a few tens of milliamps. However, to attain the required speed for modulation with load capacitances as high as 2.5 nF, the ability to both source and sink peak currents as high as 5–10 A is required. The design of a 40 kV, 250 mA high voltage amplifier has now been completed and the amplifier was fabricated. The transistors are configured into modules and the modules are mounted in four card cages within an enclosure that has a size equivalent to a standard electronics rack. The high voltage DC power supply, that provides the input voltage to the amplifier section, and the controls for the amplifier are housed within a separate electronics rack. An 800 kΩ, 2.5 nF dummy load for testing the amplifier was also fabricated, and the amplifier was been tested into this load up to its full output. The description of the recently completed amplifier and the test results will be presented. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The ITER Electron Cyclotron (EC) plant is a complex system, essential for plasma operation. The system is being designed to supply up to 20 MW of power at 170 GHz; it consists of 24 RF sources (or Gyrotrons) connected by switchable transmission lines to four upper and one equatorial launcher. The complexity of the EC plant requires a Plant Controller, which provides the functional and operational interface with CODAC and the Plasma Control System and coordinates the various Subsystem Control Units, i.e. the local controllers of power supplies, Gyrotrons, transmission lines and launchers. A conceptual design of the Electron Cyclotron Control System (ECCS) was developed, starting from the collection of the user requirements, which have then been organized as a set of operational scenarios exploiting the EC system. The design consists in a thorough functional analysis, including also protection functions, and in the development of a conceptual I&C architecture. The main aim of the work was to identify the physics requirements and to translate them into control system requirements, in order to define the interfaces within the components of the ECCS. The definition of these interfaces is urgent because some of the subsystems are already in an advanced design phase. The present paper describes both the methodology used and the resulting design. [ABSTRACT FROM AUTHOR]

China Fusion Engineering Test Reactor (CFETR) is a test tokamak which is built for magnetically confined fusion plasma experiments. The electron cyclotron (EC) wave system of CFETR is designed to inject 20 MW RF power into the plasma for heating and current drive (H&CD) applications. The EC wave system consists of RF sources, twenty transmission lines (TLs) and one equatorial launcher. RF sources contain twenty gyrotrons with the output power 1 MW. There are series of microwave components distributed along the TL and the percentage of power losses of each TL is about 8.7%. In the equatorial launcher, five RF beams are injected into one focusing mirror and then reflected to the plasma via one injection mirror. The focusing mirror is spherical to focus Gaussian beam and the injection mirror which is flat can steer in the toroidal direction. After optic calculation and optimization, all the quasi-optical parameters for launcher design are given. Combining with the thermal stress analysis, the chosen inner diameter of water channel of injection mirror is 12 mm and the suggested water velocity is 3 m/s. [ABSTRACT FROM AUTHOR]

• Work completed by the Heating and Current Drive Workpackage for DEMO during the pre-conceptual design phase (2014-2020) • The Electron Cyclotron System is the baseline for DEMO • Description of the Electron Cyclotron System • Description of the Ion Cyclotron System • Description of the Neutral Beam Heating System The European DEMO is a pulsed device with pulse length of 2 hours. The functions devoted to the heating and current drive system are: plasma breakdown, plasma ramp-up to the flat-top where fusion reactions occur, the control of the plasma during the flat-top phase, and finally the plasma ramp-down. The EU-DEMO project was in a Pre-Concept Design Phase during 2014-2020, meaning that in some cases, the design values of the device and the precise requirements from the physics point of view were not yet frozen. A total of 130 MW was considered for the all phases of the plasma: in the flat top, 30 MW is required for neoclassical tearing modes (NTM) control, 30 MW for burn control, and 70 MW for the control of thermal instability (TI), without any specific functions requested from each system, Electron Cyclotron (EC), Ion Cyclotron (IC), or Neutral Beam (NB) Injection. At the beginning of 2020, a strategic decision was taken, to consider EC as the baseline for the next phase (in 2021 and beyond). R&D on IC and NB will be risk mitigation measures. In parallel with progresses in Physics modelling, a decision point on the heating strategy will be taken by 2024. This paper describes the status of the R&D development during the period 2014-2020. It assumes that the 3 systems EC, IC and NB will be needed. For integration studies, they are assumed to be implemented at a power level of at least 50 MW. This paper describes in detail the status reached by the EC, IC and NB at the end of 2020. It will be used in the future for further development of the baseline heating method EC, and serves as starting point to further develop IC and NB in areas needed for these systems to be considered for DEMO. [ABSTRACT FROM AUTHOR]

Abstract: An upgrade of the electron cyclotron heating system on DIII-D to almost 15MW is being planned which will expand it from a system with six 1MW 110GHz gyrotrons to one with ten gyrotrons. A depressed collector 1.2MW 110GHz gyrotron is being commissioned as the seventh gyrotron. A new 117.5GHz 1.5MW depressed collector gyrotron has been designed, and the first article will be the eighth gyrotron. Two more are planned, increasing the system to ten total gyrotrons, and the existing 1MW gyrotrons will subsequently be replaced with 1.5MW gyrotrons. Communications and Power Industries completed the design of the 117.5GHz gyrotron, and are now fabricating the first article. The design was optimized for a nominal 1.5MW at a beam voltage of 105kV, collector potential depression of 30kV, and beam current of 50A, but can achieve 1.8MW at 60A. The design of the collector permits modulation above 100Hz by either the body or the cathode power supply, or both, while modulation below 100Hz must use only the cathode power supply. General Atomics is developing solid-state power supplies for this upgrade: a solid-state modulator for the cathode power supply and a linear high voltage amplifier for the body power supply. The solid-state modulator has series-connected insulated-gate bipolar transistors that are switched at a fixed frequency by a pulse-width modulation regulator to control the output voltage. The design of the linear high voltage amplifier has series-connected transistors to control the output voltage, which was successfully demonstrated in a proof-of-principle test at 2kV. The designs of complete power supplies are progressing. The design features of the 117.5GHz 1.5MW gyrotron and the solid-state cathode and body power supplies will be described and the current status and plans are presented. [Copyright &y& Elsevier]

Abstract: The Electron Cyclotron (EC) system for the ITER tokamak is designed to inject ≥20MW RF power into the plasma for Heating and Current Drive (H&CD) applications. The EC system consists of up to 26 gyrotrons (between 1 and 2MW each), the associated power supplies, 24 transmission lines and 5 launchers. The EC system has a diverse range of applications including central heating and current drive, current profile tailoring and control of plasma magneto-hydrodynamic (MHD) instabilities such as the sawtooth and neoclassical tearing modes (NTMs). This diverse range of applications requires the launchers to be capable of depositing the EC power across nearly the entire plasma cross section. This is achieved by two types of antennas: an equatorial port launcher (capable of injecting up to 20MW from the plasma axis to mid-radius) and four upper port launchers providing access from inside of mid radius to near the plasma edge. The equatorial launcher design is optimized for central heating, current drive and profile tailoring, while the upper launcher should provide a very focused and peaked current density profile to control the plasma instabilities. The overall EC system has been modified during the past 3 years taking into account the issues identified in the ITER design review from 2007 and 2008 as well as integrating new technologies. This paper will review the principal objectives of the EC system, modifications made during the past 2 years and how the design is compliant with the principal objectives. [Copyright &y& Elsevier]

Abstract: The RF heating and current drive (H&CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation. These HVPS should deliver around 50MW electrical power to each of the RF H&CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems. This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status. [Copyright &y& Elsevier]

Abstract: The electron cyclotron heating system on DIII-D has been supporting experiments with six gyrotrons. The gyrotrons are connected to three power supplies: two have single modulators, each energizing two gyrotrons, while the third has three modulators that can energize three gyrotrons asynchronously. However, only two gyrotrons can be run synchronously due to the limitation of the high voltage dc power supply providing the input voltage to the modulators. These two configurations mimic proposed architectures of power systems for multiple gyrotron systems and demonstrate their advantages or disadvantages, which will be discussed in more detail. A fourth power system is being built as part of an upgrade to eight gyrotrons. The two future gyrotrons will have depressed collectors. A 1.2MW gyrotron will be delivered in mid-2011 and the design of a 1.5MW gyrotron has been initiated. This fourth power supply will have two modulators to independently energize the cathode of each gyrotron. Commercially available high voltage amplifiers will provide the body voltage. It will also have a solid-state crowbar instead of an ignitron crowbar. This power supply is described in more detail, as well as the testing performed on the solid-state crowbar to prove its performance. [Copyright &y& Elsevier]

Abstract: A theoretical study of energy transfer by electron cyclotron resonance to tokamak plasma is presented. Then the predictions of linear theory including relativistic effects on the wave absorption are examined. Electron-cyclotron (EC) absorption in tokamak plasma is based on interaction between wave and electron cyclotron movement when the electron passes through a layer of resonance at a fixed frequency which depends on the magnetic field. This technique is the principle of additional heating (ECRH) and the generation of non-inductive current drive (ECCD) in modern fusion devices. The power absorbed depends on the optical depth which in turn depends on coefficient of absorption and the order of the excited harmonic for O-mode or X-mode [Copyright &y& Elsevier]

Abstract: The RF heating and current drive (H&CD) systems that are to be installed in ITER during the construction phase, are the electron cyclotron (EC) and ion cyclotron (IC) systems. They are complex assemblies of high voltage power supplies (HVPS), RF generators, transmission lines and antennas. Their design and integration are constrained by many interfaces, both internal, between the subsystems, and external, with the other ITER systems. In addition, some components must be compatible with a nuclear environment and are classified as Safety Important Component. This paper describes the processes implemented in ITER to ensure proper integration. [ABSTRACT FROM AUTHOR]

Abstract: The Electron Cyclotron system for ITER is an in-kind procurement shared between five parties and the total installed power will be 24MW, corresponding to a nominal injected power of 20MW to the plasma, with a possible upgrade up to 48MW (corresponding to 40MW injected). Some critical issues have been raised and changes are proposed to simplify these procurements and to facilitate the integration into ITER. The progress in the design and the integration of the EC system into the whole project is presented in this paper, as well as some issues still under studies and some recommendations made by external expert committees. [Copyright &y& Elsevier]

Abstract: The activities in Europe towards the development of the EC power sources for ITER are centered on the development of a 170GHz, 2MW, CW coaxial cavity gyrotron of collector potential depressed (CPD) type. A gyrotron with a higher unit power than the ITER reference (1MW) would yield a reduction of the installation costs, a more compact launcher design and, if required, an increase of the power delivered through one port. Tests proving the principle were successfully performed on a short-pulse experimental gyrotron delivering up to 2.2MW in single mode. Following this success, a coordinated and fully consistent programme of development has been launched. The first industrial 2MW prototype is now at an advanced stage of construction. The associated superconductive magnet producing 6.86T on the cavity axis is also being procured. Dummy loads suitable for short and CW operation are also part of the development effort. Finally, a new EC test facility, with the features necessary for the testing of the gyrotron up to full power in CW, has been established and includes a fully solid-state power supply system. [Copyright &y& Elsevier]

Abstract: We have performed one-dimensional full particle simulations of perpendicular shocks and found that an electron cyclotron microinstability can develop in the foot during the self-reformation phase of low β i supercritical shocks. The instability is excited by the beam of reflected ions interacting with the incoming electrons. It exhibits a rapid growth, and propagates along the shock normal towards upstream. This instability, which does not require high Mach number, has a frequency comparable to the electron cyclotron frequency and a wavelength shorter than the electron inertia length. It basically results from the coupling of electron Bernstein waves with an ion beam mode carried by the reflected ions. Dispersion properties in the foot are analysed. We discuss the effects of varying parameters, in particular as the fake ion-to-electron mass ratio used in the simulations converges to more realistic values. Comparison with other microinstabilities evidenced in recent full particle simulations of shocks is outlined. [Copyright &y& Elsevier]

Different systems are foreseen for the plasma additional heating and current drive of ITER-FEAT, to provide preionization, assisted current startup, bulk central heating and on- and off-axis current drive. Three different systems of radio frequency (RF) heating (ion cyclotron (IC), electron cyclotron (EC) and lower hybrid (LH)) and a couple of neutral beam injectors (NBI) were designed. The RF generators employ a number of different types of high power vacuum tubes (tetrode (IC), gyrotron (EC) and klystron (LH)) while the neutral injectors consist of coordinated ion production, acceleration and neutralization systems. All these equipments require dedicated dc power supplies PSs; their description and the discussion of their characteristics is done in the paper. * Disclaimer: This report is an account of work undertaken within the framework of the ITER EDA Agreement. The views of the author do not necessarily reflect those of the ITER Director, the Parties to the ITER EDA Agreement, or the International Atomic Energy Agency. [ABSTRACT FROM AUTHOR]

• Design validation of components for EC-assisted breakdown of ITER first plasma. • Updated design of the mirrors to reduce the beams envelope at the dump entrance. • Multi-bounces ray-tracing model developed for the design and assessment of the dump. • Modeling of the thermal load on the plates and the fraction of escaping power. • Study of effects of misalignment of beams and variations in the absorber thickness. 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. [ABSTRACT FROM AUTHOR]

The pre-conceptual layout for an electron cyclotron system (ECS) in DEMO is described. The present DEMO ECS considers only equatorial ports for both plasma heating and neoclassical tearing mode (NTM) control. This differs from ITER, where four launchers in upper oblique ports are dedicated to NTM control and one equatorial EC port for heating and current drive (H&CD) purposes as basic configuration. Rather than upper oblique ports, DEMO has upper vertical ports to allow the vertical removal of the large breeding blanket segments. While ITER is using front steering antennas for NTM control, in DEMO the antennas are recessed behind the breeding blanket and called mid-steering antennas, referred to the radially recessed position to the breeding blanket. In the DEMO pre-conceptual design phase two variants are studied to integrate the ECS in equatorial ports. The first option integrates waveguide bundles at four vertical levels inside EC port plugs with antennas with fixed and movable mid-steering mirrors that are powered by gyrotrons, operating at minimum two different multiples of the fundamental resonance frequency of the microwave output window. Alternatively, the second option integrates fixed antenna launchers connected to frequency step-tunable gyrotrons. The first variant is described in this paper, introducing the design and functional requirements, presenting the equatorial port allocation, the port plug design including its maintenance concept, the basic port cell layout, the transmission line system with diamond windows from the tokamak up to the RF building and the gyrotron sources. The ECS design studies are supported by neutronic and tokamak integration studies, quasi-optical and plasma physics studies, which will be summarized. Physics and technological gaps will be discussed and an outlook to future work will be given. [ABSTRACT FROM AUTHOR]