The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

The travelling wave array (TWA) concept was proposed as an RF actuator in the ion cyclotron range of frequencies (ICRF) for future fusion reactors and represents a relevant alternative to the conventional individually-fed arrays used in presentday machines.This paper presents an overview of the TWA concept, from the proposal for EU-DEMO to the successful tests of a high RF power mock-up paving the way for an experiment in a long-pulse device like WEST.The best TWA integration case considered for a fusion reactor like EU-DEMO would be part of the breeding blanket, sharing its cooling and effectively acting as a first-wall component.The antenna should be insensitive to considerable mechanical deformations arising from the high-temperature operation of the blanket.The roadmap to the proof-of-concept in a long-pulse device foresees a test at high power (up to 2 MW).We have designed and built a TWA antenna mock-up that has been successfully tested in the TITAN facility.This mock-up was an essential milestone that allowed us to demonstrate the validity of the design, confirming some key characteristics of the TWA concept.The ability to tune the antenna to be matched over a large frequency band was controlled in the design phase using trimmers and was well demonstrated during the experimental phase.The antenna frequency response showed to be resilient to thermo-mechanical deformations.The ohmic losses proved to be limited ( 4%).The expected operation targets were met (in TITAN) of 2 MW / 3 s and 500 kW / 60 s.In the proof-of-concept, the antenna should minimize unwanted plasma-wall interactions (PWIs).The TWA already benefits from lower fields and a very narrow power spectrum due to the larger number of straps characterizing its structure.By adjusting the layout of the antenna, we show that the power spectrum can be tuned to avoid low-|k | excitation.Furthermore, we show that the electric fields on the antenna limiters can be reduced.The status, challenges and opportunities of the WEST TWA proof-of-concept proposal are discussed and future work is outlined.

The ECRH system formerly used in Tore Supra is being upgraded to start on WEST in 2023, at a power level of 1MW and frequency of 105 GHz. Its ultimate 3MW/1000s capability is expected to enlarge the WEST operational domain by increasing margins with respect to H-mode access, and by providing additional flexibility in terms of achievable scenarios using impurity and/or MHD control. This flexibility is made possible using an antenna based on three steerable mirrors for controlled power injection. In order to determine an appropriate range of EC wave injection angles for WEST scenarios, the fast and reliable ray-tracing code REMA has been interfaced with the WEST IMAS database. This allows the EC power damping rate to be quickly assessed, as well as deposition profiles to be predicted in realistic plasma conditions. Based on a typical WEST discharge at central magnetic field B0~3.6 T, central line-averaged electron density nl~4 × 1019 m−3 and central electron temperature Te0~3keV, ray-tracing calculations have been performed. Comprehensive poloidal and toroidal angle scans, as well as variations of Bt, nl and Te0 with respect to the reference parameters have allowed an adequate range of injection angles to be determined for efficient use of ECRH and/or ECCD in typical WEST scenarios, and compared with the mechanical limits set by the antenna mechanical characteristics. In order to further characterize the effect of this new power source in WEST scenarios, EC wave deposition and current profiles from ray-tracing calculations have been included in integrated simulation codes. It has been shown that this additional power source could allow central electron heating to be achieved, potentially alleviating the issue of radiative collapse caused by impurities observed in some situations.

High power experiments, up to 9.2 MW with LHCD and ICRH, have been carried out in the full tungsten tokamak WEST. Quasi non inductive discharges have been achieved allowing to extend the plasma duration to 53 s with stationary conditions in particular with respect to tungsten contamination. Transitions to H mode are observed, and H-modes lasting up to 4 s have been obtained. The increase in stored energy is weak since the power crossing the separatrix is close to the L–H threshold. Hot L mode plasmas (central temperature exceeding 3 keV) with a confinement time following the ITER L96 scaling law are routinely obtained. The weak aspect ratio dependence of this scaling law is confirmed. Tungsten accumulation is generally not an operational issue on WEST. Difficulty of burning through tungsten can prevent the discharge from accessing to a hot core plasma in the ramp-up phase, or can lead to rapid collapse of the central temperature when radiation is enhanced by a slight decrease of the temperature. Except a few pulses post-boronization, the plasma radiation is rather high ( P rad/ P tot ∼ 50%) and is dominated by tungsten. This fraction does not vary as the RF power is ramped up and is quite similar in ICRH and/or LHCD heated plasmas. An estimate of the contribution of the RF antennas to the plasma contamination in tungsten is given.

This paper presents the current status of the WEST TWA antenna, its mock-up and a possible extrapolation to DEMO. The updated WEST TWA design has a reduced antenna length and features feeding and mechanical support from a single vessel port. A mock-up of the WEST TWA antenna was designed in 2019, manufactured during 2020 and installed in the TITAN test facility at the beginning of 2021. The results of the mock-up at low and high power, its diagnostic system and the prospects are explained. Extensions towards a TWA antenna for WEST and a possible TWA system for the future DEMO tokamak reactor are briefly discussed.

Travelling wave array (TWA) antennas have been proposed for the ICRH (Ion Cyclotron Resonance Heating) antennas of fusion reactors in view to decreasing the antenna voltage and associated electric field. This paper reports the progress of the design and structural analysis of an actively cooled high power TWA antenna for WEST. First, the design of a non-cooled mock-up that will be tested in the TITAN facility is presented. The main objective is to assess the voltage stand-off of the antenna at power and electric field levels relevant for future nominal operation in a fusion device. The main characteristics of the mock-up are detailed and the results from thermal and structural analysis of the mock-up tested in TITAN are presented. In the second part, the compatibility with WEST environment for an actively-cooled TWA antenna is assessed. Mechanical, electromagnetic, thermal and hydraulics constrains (.i.e VDE, plasma radiation, toroidal magnetic field ripple…) are listed in a system loads specification. A Virtual Reality analysis using IRFM tools has been performed to check compliance of the design in term of assembly inside the WEST vacuum vessel.

The challenge of tokamak auxiliary heating method ICRH is to couple large amount of power through the plasma boundary, where an evanescence layer has to be crossed, without exceeding the voltage standoff at the antenna. Travelling Wave Array antenna has been proposed for ICRH of future fusion reactor such as DEMO in view to decrease the antenna power density. However, so far the voltage standoff of a TWA antenna in the ICRH frequency range (30-60 MHz) has never been demonstrated. A TWA mock-up has been designed and will be tested in the CEA/IRFM testbed TITAN to assess its compatibility to RF high power operation. In this paper, the main characteristics of the TWA mock-up are reported in detail. In addition, some near-field RF analyses are presented, containing reflection characteristics, port voltage, electric field and k(//) spectrum. Meanwhile, the structural evaluation of the antenna during its lifting process is carried out to check with safety requirements. The assembly strategy of the whole structure is also explained elaborately to ensure the fluency and correct integration of the TWA mock-up. Finally, the RF assessment of antenna straps critical dimensions is briefly described.

WEST represents the ideal device to test a reactor relevant travelling wave array (TWA) due to its long pulse capability and full metal wall configuration. Moreover, the already installed ICRF high power launchers will allow a direct comparison between a classical in-port antenna and the TWA. The design and exploitation of a high power (2 MW) TWA mock-up to be tested in TITAN represent the first step towards its integration in WEST. This paper presents the progress in the RF characterization of this high power mock-up. An integrated modelling scheme based on the full-wave code HFSS and on Python packages for RF circuit simulations is developed and used to parametrically analyse the response of the mock-up inside TITAN. The sensitivity of the antenna to geometrical deformations, e.g. thermal dilatation or manufacturing tolerances, is analysed. The use of trimmers to compensate those deformations is discussed and results from a manufactured trimmer test-bed are presented. A diagnostic system is proposed to measure electric field and current on the straps. Aspects like sensitivity and calibration are briefly discussed. In conclusion, the next steps of the project are outlined.

Three identical new WEST ion cyclotron resonance heating (ICRH) antennas have been designed, assembled then commissioned on plasma from 2013 to 2019. The WEST ICRH system is both load-resilient and compatible with long-pulse operations. The three antennas have been successfully operated together on plasma in 2019 and 2020, with up to 5.8 MW of coupled power. The load resilience capability has been demonstrated and the antenna feedback controls for phase and matching have been developed. The breakdown detection systems have been validated and successfully protected the antennas. The use of ICRH in combination with lower hybrid has triggered the first high confinement mode transitions identified on WEST.

As part of ITPA-Integrated Operational Scenario activities, this contribution reviews recent experimental characterizations of radio-frequency (RF)-induced scrape-off layer (SOL) modifications of various tokamaks worldwide and of the LArge Plasma Device at UCLA. The phenomenology, as observed using a large variety of measurement techniques, is consistent with expectations from RF sheath rectification. Emphasis is then put on the complex three-dimensional (3D) spatial patterns of the RF–SOL interaction, in relation to the magnetic topology and the spatial distribution of RF currents over the metallic structures surrounding the RF wave launchers. Dependence on the local plasma parameters in the antenna vicinity is also briefly addressed. The final part discusses implications for future devices.

In 2018, Ion Cyclotron Resonant Frequency (ICRF) waves were for the first time applied to the WEST plasma scenarios. In ICRF-only plasmas at medium density, or on top of a low level of Lower Hybrid (LH) power, the coupled ICRF power increases the plasma energy content. However in discharges with large LH power at high core density, nearly all the applied ICRF power gets radiated, mainly in the plasma bulk. Both the energy content and conducted power decrease. Two peculiarities of WEST, that may combine, are presently invoked to explain this phenomenology:1) Fast ion ripple losses, evidenced between TF coils on the baffle, likely degrade the ICRF heating efficiency. 2) RF-enhanced W sources, evidenced on several Plasma-Facing Components (PFCs), likely over-contaminate the high-power plasmas.

The paper discusses the characterization of the three high power steady-state and load-resilient ICRF launchers of WEST before their installation in the tokamak. These launchers have been characterized and validated in low-power experiments (milliwatt range) as well as in experiments at the nominal RF voltages and currents in the TITAN vacuum chamber (similar to 30 kV and 915 A peak). The successful commissioning of two of the launchers during the WEST C3 campaign at similar to 1 MW power level is illustrated. Manual and real-time controlled impedance-matching of the launchers are discussed, as well as the validation of their load-resilience. Furthermore, several redundant and complementary protection systems have been validated and are reviewed in the paper.

This study compares experimental observations on the two challenges of Ion Cyclotron Resonance Heating (ICRH), namely the coupling of waves to the plasma and the enhanced impurity sputtering in EAST and WEST medium size tokamaks. In WEST, Lower Hybrid (LH) power helps improving ICRH coupling. In both machines experiments reveal that fueling from the midplane not only helps to couple waves from nearby antennas like in other devices, but also has an impact on the scrape-off layer (SOL) density in regions that are not magnetically connected to the valves. Localized midplane nozzle valves allow similar (in WEST) or better (in EAST) coupling compared to poloidally distributed valves. Core density control requirements for long-pulse operation, in particular in L-mode regime, however limit the amount of gas that can be injected. During ICRH impurities can contaminate the plasma up to a level detrimental for the operation, e.g. up to 100% of ICRH power can be radiated on WEST in certain high power conditions. In WEST, tungsten (W) production measured by visible spectroscopy increases on all the observed objects during ICRH compared to a reference phase without ICRH. On some components (antenna side limiters, baffle, divertor) the rise is larger than with a similar LH power. The relative contribution of each object and physical process (RF-sheaths, fast ion ripple losses) to core contamination yet remains poorly known. Comparing antenna limiters with W-coating vs low-Z materials would help quantifying the role of these components. In EAST, the core W content, measured by EUV spectroscopy in presence of divertor sources only, is correlated with the total injected power, either from ICRH or LH. Since 2018 the LH guard limiter tiles were W-coated. Their contribution to the core W content appears more important than divertor sources when two-strap ICRH antennas magnetically connected to W components at the midplane is powered, compromising high performance operations.

Travelling wave array (TWA) antennas distributed along the periphery of the tokamak are presently considered as an ion cyclotron resonance frequencies (ICRF) heating solution for the DEMO reactor. Compared to the conventional ICRF antenna systems currently in use or designed for future machines like ITER, the TWA consists of antenna sections integrated in the breeding blanket scattered around the machine, each one fed through a variable coupler in a resonant ring configuration. Previous modelling of an antenna system for DEMO with 16 quadruple TWA sections of eight straps shows that a power capability exceeding 50 MW can be obtained in the frequency band of interest using the reference low coupling plasma profile of ITER. The described system optimizes the coupling to the plasma by providing a large number of radiating elements, which results in enhanced antenna directivity, hereby decreasing the antenna power density. This results in a maximum strap voltage amplitude of only 15 kV and maximum inter-strap voltage amplitude of 18 kV. The generators remain matched for all loading conditions: the system is totally load resilient. Following the recommendation of the work package heating and current drive Review Panel, a TWA ion cyclotron resonant heating (ICRH) system consisting of fewer sections concentrated in front of the equatorial ports is analysed in this paper and compared to the previous design. Reducing the number of sections increases the power density and its associated voltages. To couple 50 MW on the ITER density profile, voltages up to 30 kV are now required. Some aspects like the coupling between sections and its repercussion on the feeding network are briefly discussed. To assess the feasibility of the TWA fed by a resonant ring as an ICRH system for a DEMO reactor, a test on an existing medium size tokamak is under study.

To decrease the power density and associated high voltages, a distributed antenna system is proposed as ICRH system for the DEMO reactor. Among the different solutions, a layout made from a set of travelling wave array (TWA) sections is considered as the most promising. It optimizes coupling to the plasma, is load resilient and avoids large values for the VSWR in the feeding lines. The total radiated power scales as the number of independently fed sections such that high reliability can be expected. The TWA concept for ICRH is innovative and very different from the traditional IC antennas. A test on WEST would provide a proof of principle of the validity of the TWA approach together with a comparison with the existing WEST IC antennas. The chosen geometry of the TWA section is compatible with one unit of the complete set designed for a future reactor. The paper describes the progress made in the preparation of a test on WEST along with the extrapolation for a future reactor like DEMO. A comparative modeling with the present antennas is also discussed and a preliminary RAMI analysis is introduced showing the promising positive impact of the TWA design on the RAMI scores.

Introduction: The WEST tokamak has achieved L-mode X-point plasmas with 12 s duration (IP = 0.5 MA, ne ~ 3×10 19 m -3 , BT = 3.7 T) during the last experimental campaign in summer 2018 [1]. WEST is the first and only full W-device relying on radiofrequency systems for heating and current drive. This paper presents the results of the Lower Hybrid Current Drive (LHCD) experiments in WEST, together with the commissioning and successful test on plasma of the first new load-resilient long-pulse Ion Cyclotron Range of Frequencies (ICRF) antenna, manufactured in collaboration with ASIPP, Hefei [2]. LHCD results: The high power CW LHCD system (9 MW at generator, two launchers, f = 3.7 GHz) has been extensively used in Tore Supra mainly for long pulse operation, and has since undergone modifications and maintenance for WEST. In the first WEST experimental campaign (Oct. 2017 – Feb. 2018), several plasma equilibria were tested aiming at optimising the coupling of the Lower Hybrid (LH) power. Once adequate plasma equilibria had been found [3], the LH power could rapidly be increased to 2.3 MW for 2 s [4]. In the summer campaign 2018, two boronisations were carried out, which allowed to operate at higher plasma density (ne ~ 3×10 19 m -3 ), and thus to improving the LH coupling further. L-mode discharges lasting up to 12 s with coupled LH power of 1.9 MW for 9.5 s was achieved (Fig. 1), with low level of reflected power (reflection coefficient < 8% on both launchers). The LH current drive efficiency is found to be similar to that obtained in Tore Supra GJ discharges ( ~ 0.7×10 19 m -2 A/W) [5]. The improved current drive efficiency as compared to the first WEST experimental campaign [4] could partly be due to lower Zeff after the boronisations.

Rhodium (Rh) coating on CuCrZr substrate is a promising material option for optical, structural and electrical applications on nuclear fusion reactors. For these applications, Rh coated CuCrZr components subject to long time of thermal ageing due to pre-treatment or normal operation condition. In this paper, both finite element method (FEM) and experimental method were applied to investigate the effects of thermal ageing on mechanical performance of Rh coating after 250 degrees C, 500 h baking in vacuum. Based on FEM analysis, thermal stresses which concentrate at Rh coating interface is the main source of cracking, and such stresses can be minimized efficiently by introducing a 0.5 mu m Au interlayer into the coating layer structure. According to thermal ageing experiments, through-thickness cracking in the Rh coating due to thermal stress releasing and voids generated at the Rh bonding interface caused by Kirkendall effect were the main micro-structure changes in the coating system. The solid-solution hardening caused by significant Cu diffusion into Rh is the dominant factor that affected the Rh coating's hardness. The existing of large amount of cracks in the Rh coating and voids at the Rh coating interface deteriorated the adhesion performance of Rh on CuCrZr substrate by 30%.

The ITER Ion Cyclotron Resonance Heating (ICRH) system is designed to couple to the plasma 20 MW of RF power from two antennas in the 40-55 MHz frequency range during long pulses of up to 3600 s and under various plasma conditions with Edge Localized Modes. Radio-Frequency (RF) contacts are integrated within the ITER ICRH launcher in order to ensure the RF current continuity and ease the mechanical assembly by allowing the free thermal expansion of the Removable Vacuum Transmission Line coaxial conductors during RF operations or during 250 degrees C baking phases. A material study has been carried out to determine which materials and associated coatings are relevant for RF contacts application in ITER. In parallel, RF tests have been performed with a new prototype of Multi-Contact (R) LA-CUT/0,25/0 contacts made of silver-coated CuCrZr louvers. During these tests on a RF vacuum resonator, currents between 1.2 kA and 1.3 kA peak have been reached a few tens of times in steady-state conditions without any visible damage on the louvers. A final 62 MHz pulse ending in a 300 s flat top at 1.9 kA resulted in severe damage to the contact. In addition, a test bed which performs sliding test cycles has been built in order to reproduce the wear of the contact prototype after 30 000 sliding cycles on a 3 mm stroke at 175 degrees C under vacuum. The silver coating of the louvers is removed after approximately a hundred cycles whilst, to the contrary, damage to the CuCrZr louvers is relatively low.

Radio-frequency (RF) contacts-which are an example of electrical contacts-are commonly employed on accelerators and nuclear fusion experimental devices. RF contacts with a current load of 2 kA for steady-state operation were designed for application to the International Thermonuclear Experimental Reactor (ITER) device. In contrast to the typical working conditions of general commercial electrical contacts, those of RF contacts employed on fusion devices include high vacuum, high temperature, and neutron radiation. CuCrZr is currently of interest as a base material for the manufacture of louvers of RF contacts, which has excellent thermal and electrical properties and has low creep rate at 250 degrees C. In this study, a hard Au coating (Au-Ni) was electroplated on CuCrZr samples and the samples were then subjected to thermal aging treatment at 250 degrees C for 500 h in order to simulate the vacuum-commissioning process of the ITER. The effects of thermal aging on the hardness, elastic modulus, crystallite size, and compositions of the coating were investigated via microstructural and mechanical characterizations of the coating material. Metal atom migration in different coating layers during thermal aging was characterized and evaluated via scanning electron microscopy/energy dispersive X-ray spectroscopy observations of the cross-sectional surfaces, and the obtained results could be used to directly select the coating thickness for the final RF contact component. The contact resistance-an important parameter of the RF contact- was measured in a dedicated testbed built to simulate fusion reactor conditions between CuCrZr pins and stainless steel plates coated with Au-Ni and Rh, respectively.