Preliminary design of the ITER ECH Upper Launcher

The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q = 1 (sawtooth instability), q = 3/2 and q = 2 (NTMs).In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design     

Operation of an ITER relevant inspection robot on Tore Supra tokamak

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. CEA has developed a multipurpose carrier able to realize deployments in the plasma vessel without breaking the Ultra High Vacuum (UHV) and temperature conditioning. A 6 years R&D programme was jointly conducted by CEA-LIST Interactive Robotics Unit and the Institute for Magnetic Fusion Research (IRFM) in order to demonstrate the feasibility and reliability of an in-vessel inspection robot relevant to ITER requirements.The Articulated Inspection Arm robot (AIA) is an 8-m long multilink carrier with a payload up to 10 kg operable between plasma under tokamak conditioning environment; its geometry allows a complete close inspection of Plasma Facing Components (PFCs) of the Tore Supra vessel.Different tools are being developed by CEA to be plugged at the front head of the carrier. The diagnostic presently in operation consists in a viewing system offering accurate visual inspection of PFCs. Leak detection of first wall based on helium sniffing and laser compact system for carbon co-deposited layers characterizations or treatments are also considered for demonstration.In April 2008, the AIA robot equipped with its vision diagnostic has realized a complete deployment into Tore Supra and the first closed inspection of the vessel under UHV conditions. During the upcoming experimental campaign, the same operation will be performed under relevant conditions (10^?6 Pa and 120 °C) after a conditioning phase at 200 °C to avoid outgassing pollution of the chamber.This paper describes the different steps of the project development, robot capabilities with the present operations conducted on Tore Supra and future requirements for making the robot a tool for tokamak routine operation.     

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current I_p ≤ 1 MA, toroidal field B_T ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s.     

Laser-assisted cleaning of beryllium-containing mirror samples from JET and PISCES-B

A set of seven polycrystalline mirror samples retrieved from the JET tokamak has been cleaned in vacuum using a pulsed laser system. The surfaces of samples exposed to plasma during 2008–2009 campaigns as part of the second phase of a comprehensive first mirror test contained a mixture of carbon, beryllium and tritium. For this reason, the samples were treated in a vacuum chamber constructed specially for this purpose. In some cases mirrors show an increase of the specular reflectivity after cleaning, though beryllium and carbon deposits were not fully removed. Additionally, three samples coated in PISCES-B with a 110–120 nm beryllium layer were subjected to laser cleaning tests as well.     

Investigations of single crystal and polycrystalline metal mirrors under erosion conditions in TEXTOR

Metal mirrors are planned for optical diagnostic systems of ITER. However, erosion, deposition and particle implantation can change the performance of mirrors. Mirrors made from the single crystal (SC) materials are among the main candidates for use in ITER diagnostic systems operating under erosion-dominated conditions. Laboratory tests have confirmed good optical performance of SC mirrors under erosion, but the dedicated direct comparative test in tokamak environment was missing.Such a direct test was performed in TEXTOR. Single crystal molybdenum, tungsten and polycrystalline (PC) molybdenum mirrors were exposed under the same conditions in the SOL plasma of TEXTOR. Surface and optical properties of mirrors were characterized before and after exposure. Before exposure glow discharge cleaning in hydrogen restored the reflectivity of mirrors oxidized during storage on air.No significant changes in total reflectivity were observed for all mirrors after exposure. Drastic increase of diffuse reflectivity was measured for PC Mo mirror, no change for a SC one. Thus, specular reflectivity of single crystal is higher than of polycrystalline one. The most affected wavelength range is 250-1000 nm, no significant change of reflectivity was noticed in the range 1000-2000 nm. Negligible effect of the exposure on polarization characteristics was observed.     

Design study of a wide-angle infrared thermography and visible observation diagnostic on JT-60SA

Design study of a wide-angle infrared (IR) thermography (surface temperature measurement) and visible observation diagnostics for JT-60SA are reported. The new design offers an optical solution without a “blind spot” which is one of the advantages. In order to image a large section inside the vacuum vessel (both in poloidal and toroidal directions), the optical system of endoscope is to provide a wide-angle view in the IR and visible wavelength ranges. The estimated IR optical spatial resolution is approximately 2 cm at a distance of 7.6 m from the front optics with a pupil diameter of 4 mm. For a surface temperature measurement it would be larger (∼4 cm for a surface temperature error less than 5%). The optics of this system can be divided into three parts: (1) a mirror based optical head (two set of spherical mirrors plus two flat mirrors) that produces an intermediate image, (2) a Cassegrain telescope system, and (3) a relay group of lenses, being adapted to the two kinds of detectors for IR and visible observations.     

Overview of engineering design,manufacturing and assembly of JT-60SA machine

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R&D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.     

First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.     

Theoretical and experimental studies on molybdenum and stainless steel mirrors cleaning by high repetition rate laser beam

Our studies were aimed to determine the damage threshold of molybdenum (Mo) and stainless steel (SS) mirrors to provide the maximum fluence which the mirror surfaces could withstand without affecting their reflectivity properties. A high repetition rate ytterbium fiber laser (20 kHz, 1.06 μm, 120 ns) was applied. The experimental single-pulse and multiple-pulse damage thresholds were obtained. To calculate damage thresholds, a 1D analytical model which takes into account the temperature dependent absorptance and multiple-pulse damage based on plastic deformations accumulation was applied. The experimental damage thresholds and the theoretical ones are in a good agreement. Cleaning tests with the contaminated mirrors exposed in JET have been performed.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

Nanostructured rhodium films for advanced mirrors produced by Pulsed Laser Deposition

In this paper advantages in the production by Pulsed Laser Deposition (PLD) of nanostructured, nanoengineered rhodium films to be used in tokamak First Mirrors (FMs) are shown. The peculiar PLD capability to tailor film structure at the nanoscale gives the possibility to deposit low roughness Rh films with a wide variety of structures and morphologies. By a proper movimentation of the substrate and using high fluence (10–19 J/cm²) infrared laser pulses, it has been possible to deposit planar and homogeneous Rh films effectively suppressing surface defects on areas of the order of 10 cm² with a satisfactory specular reflectivity. Multilayer deposition has been exploited to produce coatings with high adhesion and good mechanical properties. Finally, an estimation of the requirements to produce by PLD rhodium films suitable for the requests of ITER is provided.     

The JET programme in support of ITER

As part of its mission to prepare the operation of ITER, a major programme of enhancements has just been completed on the JET tokamak. These enhancements include a complete replacement of the plasma-facing components in JET, from carbon-based to the combination of beryllium and tungsten foreseen for ITER, an upgrade of the neutral beam heating available on JET from 20 MW/short pulse to 30 MW/long pulse operation, the installation of a high frequency pellet injection system for plasma fuelling and ELM control studies, an upgrade to the JET vertical stability system and a suite of new diagnostics.The future JET programme is foreseen to proceed progressively from a test of fuel retention in the standard regimes of ITER operation towards more aggressive, high performance experiments that will demonstrate the operating space limits with the new wall. Depending on the results of the earlier experiments, the exploitation of the enhancements is foreseen to be completed with a deuterium-tritium experiment. This would represent the most integrated test of ITER operational scenarios possible before ITER itself.JET is a cooperative programme funded and exploited in collaboration by all of the European fusion laboratories. As such, JET is a test bed for multi-national use of a single fusion facility, as is foreseen for ITER. Opportunities for broadening the participation in JET to other ITER Parties are presently being explored. If these opportunities can be implemented, JET would provide not only an integrated test of ITER regimes of operation but also a demonstration of how ITER will be operated, even to the extent of including significant numbers of the same team who will eventually operate ITER.     

Overview of EU DEMO design and R&D activities

One important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several 100 MW of net electricity to the grid and operating with a closed fuel-cycle by 2050. This is currently viewed by many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. This paper outlines the DEMO design and R&D approach that is being adopted in Europe and presents some of the preliminary design options that are under evaluation as well as the most urgent R&D work that is expected to be launched in the near-future. The R&D on materials for a near-term DEMO is discussed in detail elsewhere.     

Substrate-dependent deposition efficiency on mirrors exposed in the TCV divertor

Given the planned extensive use of metallic mirrors in the optical diagnostics systems of ITER, the study of reflectivity changes induced by erosion and/or redeposition of impurities on the mirror surfaces is of primary importance for the reliability of the diagnostics signals. This contribution will demonstrate that the mirror material choice can exert a significant influence on the relative importance of erosion/redeposition affecting the mirror reflectivity. A dedicated manipulator has been designed to allow exposure of mirror samples in the divertor region of the Tokamak à Configuration Variable (TCV) tokamak. Mirrors from different materials have been exposed both during short experimental campaigns and boronisation procedures. Before and after exposures the mirrors were characterized with different surface analysis techniques (XPS, SIMS, SEM, EDX, ellipsometry). Under identical exposure conditions, the mirror material can strongly influence the deposit thickness found on the sample: the carbon layer thickness on a Si sample is much higher than on a Mo sample. These results have potentially important consequences for the first mirror material choice in ITER.     

Overview of magnetic control in ITER

ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination.Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios.Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design.Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the plasma boundary), and the strong component of eddy current at the start-up (resulting in a poor S/N ratio for plasma reconstruction for I_p < 2 MA requiring a robust plasma control) make the ITER magnetic diagnostic system a demanding part of the magnetic control and investment protection systems. Finally the paper will illustrate the identified roles of magnetic control in the PCS (plasma control system) as formally defined in the recent first step of the design and development of the system.     

Recent and future upgrades to the DIII-D tokamak

Research on the DIII-D tokamak focuses on support for next-generation devices such as ITER by providing physics solutions to key issues and advancing the fundamental understanding of fusion plasmas. To support this goal, the DIII-D facility is planning a number of upgrades that will allow improved plasma heating, control, and diagnostic measurement capabilities. The neutral beam system has recently added an eighth ion source and one of the beamlines is currently being rebuilt to allow injection of 5 MW of off-axis power at an angle of up to 16.5° from the horizontal. The electron cyclotron heating (ECH) system is adding two additional gyrotrons and is using new launchers that can be aimed poloidally in real-time by an improved plasma control system. The fast wave heating system is being upgraded to allow two of the three launchers to inject up to 2 MW each in future experiments. Several diagnostics are being added or upgraded to more thoroughly study fluctuations, fast ions, heat flux to the walls, plasma flows, rotation, and details of the plasma density and temperature profiles.     

Thermal and mechanical analysis of ITER-relevant LHCD antenna elements

A 20 MW Lower Hybrid Current Drive system using an antenna based on the Passive-Active Multijunction (PAM) concept is envisaged on ITER. This paper gives an overview of the mechanical analysis, modeling and design carried out on two major elements of the antenna: the grill front face, and the RF feed-through or windows. The front face will have to withstand high heat and fast neutrons fluxes directly from the plasma. It will be actively cooled and present a beryllium coating upon ITER requirement. The RF window being a critical safety importance class component (SIC) because of its tritium confinement function, two of them will be put in series on each line to achieve a double barrier. A design of a water cooled 5 GHz CW RF “pillbox” window capable of sustaining 500 kW of transmitted power is proposed. Both studies allow to move forward, and focus on critical issues, such as manufacturing processes and R&D associated programs including tests of mock-ups.     

Study of large gap 200 kV breakdowns in vacuum with high stored energy for ITER neutral beam accelerator

In the context of the ITER contract “ITER/CT/07/219–200 kV Stored Energy Tests”, electrical breakdown tests have been performed in vacuum with a stored energy of up to 425 J. The experiments have been conceived and performed with the collaboration of Consorzio RFX. The tests are being performed in the 1 MV test facility at IRFM, CEA-Cadarache. They should simulate the conditions that will be found in the ITER Neutral Beam accelerator, at 200 kV. This paper presents the set-up of the test bed, the choice of critical components, the diagnostic equipments and the results obtained with 200 kV applied on the anode electrode.     

Spectroscopic studies of homogeneous thin carbon erosion films on mirrors and flakes with a high deuterium content formed in tokamak T-10

Redeposited hydrocarbon films on plasma facing elements in tokamaks accumulate hydrogen isotopes. In the present study such films were made to redeposit on stainless steel mirror substrates as thin films and without any substrate as bare flakes with high deuterium content, under deuterium-plasma discharges inside T-10 tokamak vacuum chamber. These films were subjected to spectral characterizations through Fourier-transform infrared (FT-IR), electron paramagnetic resonance (EPR), and photoluminescence techniques. IR spectra showed the presence of two main deuterium states as observed by the CD_2,3 sp³ stretching modes at 2100–2200 cm⁻¹ and the CD₂ sp³ bending modes at 600–1100 cm⁻¹. Among these, CD₃ stretching mode at 2217 cm⁻¹ may serve as a control for deuterium desorption during the cleanup process of the reactor. As a comparative measure, C60 films were also studied, the luminescence excitation spectrum of which showed similarity in peak positions with tokamak bare flakes pertained to sp² luminescence centers. The observed spectral differences are mainly due to more localized sp² states for C60 and sp³ states for tokamak flakes. EPR spectra of the bare flakes showed the defective states with a high spin density, ∼10¹⁹ cm⁻³ which serve as luminescence quenching centers, and provide a path for hydrogen isotopes adsorption.