W environment in steady-state tokamak (WEST) experiments aim to achieve long plasma pulses (1000 s) and to expose ITER-like tungsten divertor to power fluxes up to 10 MW/m2. To increase the margin to reach H-Mode regimes and control W-impurities in the plasma, the WEST ECRH system is upgraded to a power capability of 3 MW/1000 s at the frequency of 105 GHz. The Tore Supra ECRH antenna has first been improved and is now installed in the equatorial plane of WEST to start ECRH operation on plasma. On the power generation side, a new gyrotron (1 MW/1000 s) designed jointly by KIT and NKUA and manufactured by THALES has been tested in the FULGOR test bench at KIT. The validation of the gyrotron electron optic and Radio Frequency (RF) design with performances of 1 MW/5 s triggered the assembly of the second and third series gyrotrons. In parallel to the gyrotron validation and the installation of the upgraded antenna in the tokamak, the ECRH transmitter has undergone profound changes to adapt the auxiliaries to the new components (gyrotron, superconducting magnet, RF load) such as the cooling system, the high voltage power supply (HVPS) system and the plant environment itself. To control, monitor and ensure the safety of staff and components, the control system, software and communication protocols with the auxiliaries of the gyrotron have been completely refurbished. The commissioning of the complete system now equipped with the first gyrotron started in September 2024 in WEST with the goal of achieving the first ECRH shots during plasma experiments in April 2025. This article reports the main results obtained with the first gyrotron, focuses on the upgrades of the ECRH plant and its commissioning and finally gives a summary of the first experiments with ECRH in WEST.

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 aim of the WEST experiments is to master long plasma pulses (1000s) and expose ITER-like tungsten wall to deposited heat fluxes up to 10 MW/m2. To increase the margin to reach the H-Mode and to control W-impurities in the plasma, the installation of an upgraded ECRH heating system, with a gyrotron performance of 1MW/1000s per unit, is planned in 2023. With the modifications of Tore Supra to WEST, simulations at a magnetic field B0-3.7T and a central density ne0-6 x 1019 m- 3 show that the optimal frequency for central absorption is 105 GHz. For this purpose, a 105 GHz/1MW gyrotron (TH1511) has been designed at KIT in 2021, based on the technological design of the 140 GHz/1.5 MW (TH1507U) gyrotron for W7-X. Currently, three units are under fabrication at THALES. In the first phase of the project, some of the previous Tore Supra Electron Cyclotron (EC) system components will be re-installed and re-used whenever possible. This paper describes the studies performed to adapt the new ECRH system to 105 GHz and the status of the modifications necessary to re-start the system with a challenging schedule.

The French tokamak WEST supports the ITER design and operation. IRFM is designing a new Thomson scattering diagnostic to measure plasma density and temperature profiles. The diagnostic system consists of an endoscope inside a vacuum vessel, composed of actively cooled optical components. In order to validate and guarantee the diagnostic performances during normal operations, mechanical, thermal, hydraulic and vibratory behavior must be checked. Moreover, perpendicular displacement of the optical surface shall not be higher than 40 µm. Since this diagnostic operates in near infrared light, the temperature of all components must stay lower than 200 °C as not to bias the measurements. The differences in water temperature and pressure between the inlet and outlet of the diagnostic must be lower than 50 °C and 5.6 bar, respectively. The natural frequencies of the structure must be higher than 20 Hz and far enough from the frequency of external components. In this study, the worst radiative plasma scenario was chosen. The results of this study validate the accuracy of the measurements. Before manufacturing, electromagnetic disruption events must also be considered.

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017-2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m(-2) were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

The equatorial visible and infrared Wide Angle Viewing System (WAVS) for ITER is one of the key diagnostics for machine protection, plasma control and physics analysis. To achieve these objectives, the WAVS will monitor the surface temperature of the Plasma Facing Components (PFCs) by infrared (IR) thermography (3-5 mu m range) and will image the edge plasma emission in the visible range. It will be composed of 15 lines of sight installed in four equatorial ports (no. 3, 9, 12 and 17) in order to survey at least 80% of the overall area of the vacuum vessel. This paper presents the optical design of port-plug which has to cope with both challenging performances (wide field of view, wide temperature range 200-2000 degrees C, high measurement accuracy required lower than 20%) and severe constraints such as harsh nuclear environment, complex interfaces. Indeed the optical components are embedded within the ITER Diagnostic Shielding Module, DSM, which is not usual for diffraction limited instruments. It is shown that nominal optical performance of the Port-Plug are limited by the diffraction only. In operation, this design has demonstrated its superiority on the former ones since it is much less sensitive to tolerance variations.

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the TER tungsten divertor strategy. The WEST project aims to test W monoblock Plasma Facing Units (PFU) under long plasma discharge (up to 1000 s), with thermal loads of the same magnitude as those expected for ITER. Therefore the divertor is a key component of the WEST project, and so is its support structure, which has to handle strong mechanical loads. The WEST upper and lower divertor are made of 12 30 sectors, each one composed of 38 PFU that can be made of tungsten, CuCrZr or graphite. A generic 316L stainless steel 30 conic support plate is used to hold the 38 PFU together, regardless of their material. The PFUs are fixed on the support plate thanks to 152 Xm19 stainless steel fixing elements (4 per PFU), and in each of this fixing element an Aluminium-Nickel-Bronze alloy (Al-Ni-Br) pin is engaged in a slotted hole, in order to allow thermal expansion in the length direction of the PFU. The support plate is fixed on the divertor coil casing thanks to 10 M10 screws. Mechanicals loads which act on the PFUs are transmitted to the support plate through the fixing elements. These loads are due to Vertical Displacement Event (VDE), disruptions and thermal expansion of the PFU. First the different load cases, PFU configurations and scenario are presented. Then an ANSYS plastic mechanical simulation is performed in order to validate the number of cycles of the support plate for each scenario: 30,000 cycles in steady-state and 3000 cycles in VDE. Finally reactions forces from the previous ANSYS simulation are used in order to calculate the stress in the M10 screws. (C) 2017 Elsevier B.V. All rights reserved.

In order to fully validate "ITER-like" actively water cooled tungsten plasma facing units, addressing the issues of long plasma discharges, an axisymmetric divertor structure has been studied and manufactured for the implementation in the WEST (W-Tungsten Environment in Steady state Tokamak) tokamak platform. This assembly, called divertor structure and coils (4 m diameter, 20 t), is composed of two stainless steel casings containing an actively water cooled (up to 180 degrees C, 4 MPa) copper winding pack designed for a conductor current in the range of 12.5kA (up to 1000s). It must sustain harsh environmental conditions in terms of ultra-high vacuum, high temperatures and electrodynamic loads. One major difficulty is the assembly by induction brazing of individual bended conductor sectors inside the vacuum vessel and the consecutive sealing of the casings by TIG welding. Therefor development activities have been carried out on a scale one dummy coil, such as brazing, assembly, thermal cycling and electrical insulation tests (5 kV ground voltage). Whereas the brazing assembly technics and the conductor installation were validated without major difficulties, different technical solutions for the electrical insulation had to be tested. The chosen solution is a resin epoxy impregnated fiber glass fabric layered around the conductors followed by a polymerization procedure. In parallel the manufacturing of divertor structure components started in the second half of 2013 with a total delivery at the end of 2015. The paper will illustrate the technical developments which have been performed in order to fully validate the design. It concerns mainly the dummy coil and the complex conductor installation procedureassisted by virtual reality tools. The manufacturing methods proposed by industry in order to fulfil the technical requirements will be also addressed. Finally the processes and associated tools used in order to implement this large component inside the WEST vacuum vessel will be detailed. (C) 2017 Elsevier B.V. All rights reserved.

The WEST (Tungsten (W) Environment in Steady-state Tokamak) is an upgrade of Tore-Supra (TS) which aims it into an X-point magnetic configuration tokamak equipped with an actively cooled tungsten divertor. To be a platform of ITER technologies of high heat flux components testing, three sets of Ion Cyclotron Resonant Heating (ICRH) antennas have been designed to inject 9 MW during 30s or 3 MW during 1000s.The antenna design is based on a load resilient prototype successfully tested in Tore Supra in 2007. In order to allow continuous-wave (ON) operations, the mechanical design of the WEST ICRH antenna is emphasized on its cooling performances by designing fully active cooling structure. Two kinds of cooling water loops are used, with temperature and pressure of 70 degrees C/30 bar and 25 degrees C/5.2 bar, respectively. The hot water loop is used for the Faraday screen (FS) and the housing box (HB), while the cold water loop is used for the straps, the matching capacitors and the impedance transformer. To enhance the heat removal ability and control the pressure drop, the cooling channels in the FS and HB are drilled directly and parallel connected as much as possible.By performing the hydraulic thermal analysis, the lack of cooling efficiency was found in the front face of lateral collector where 1 MW/m(2) is imposed and fluid dead zones were found in some of the bars. After optimization, the cooling performance of the cooling circuit increased significantly. With a mass flow rate of 2.5 kg/s, the total pressure drop is 3.1 bar, and the peak temperatures on the FS and HB are 500 degrees C and 261 degrees C, respectively. Besides, no cavitation is found in the total cooling circuit. (C) 2015 Elsevier B.V. All rights reserved.

The WEST project (Tungsten (W) Environment in Steady state Tokamak) is an upgrade of the existing fusion machine, Tore Supra. The goal is to equip the tokamak with a fully cooled tungsten divertor and to transform the machine in a test platform open to all ITER partners.'The main assembly challenge of this project consists of an implementation of two magnet systems, called divertors, with an accuracy of 1 mm. Indeed, each divertor has about 4m as diameter and has a heavy weight of 10 tons; also it introduces piece by piece in the original vessel through tight ports then assembled inside. To ensure a perfect fitting between these new components and a very constrained environment, it is necessary to use the latest CAD technologies available.Beyond conventional CAD tools, the virtual reality (VR) room of the institute provides several useful tools. Thanks to the 185" stereoscopic 3D screen and a force feedback arm linked to clash detection software developed by the CEA LIST, a new way to carry out design and assembly studies was performed. In order to improve VR results, metrology data (3D scan) enhance simulations. Therefore, it becomes possible to be aware of the real size of a component and future difficulties in assembling it.At last, performance of such simulations is evaluated and compared to physical mockup in order to bring enhancement to the VR tools, before to be compared to the real operations on Tore Supra. The aim is to build a design tool that helps the designer since early stage of the design of complex systems, taking into consideration integration, assembly and maintenance aspects while reducing costs and schedule of a project. (C) 2015 Elsevier B.V. All rights reserved.

In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m(2) heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design.One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30 degrees sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30 degrees sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected. (C) 2015 Elsevier B.V. All rights reserved.

In order to fully validate "ITER-like" actively water cooled tungsten plasma facing units, the implementation of an axisymmetric divertor structure in the Tokamak Tore-Supra has been studied. With this major upgrade, the so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the issues of long plasma discharges using a tungsten divertor based on monoblock targets.The divertor structure and coils assembly are made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 degrees C, 4 MPa) in which a total divertor current of up to 16 x 13 kA is circulating in steady state. The conductor is electrically insulated and wedged inside the casing in order to be mechanically protected.The divertor which is designed to perform steady state plasma operation (up to 1000 s), must sustain harsh environmental conditions in terms of ultra light vacuum conditions, electromagnetical loads and electrical insulation (5 kV ground voltage) under high temperature (180 degrees C).Therefore, a feasibility study of such a complex structure has been performed. It implied activities on a scale one dummy coil, such as installation, assembly issues and representative tests (electric, thermal and hydraulic).The manufacturing of the divertor structure, which is a large assembly of 4-m diameter representing a total weight of around 20 tonnes, started in the second half of 2013 and is expected to be delivered by the end of 2014.The paper will illustrate the technical developments and tests performed during 2013 and beginning of 2014 in order to fully validate the design concept before the industrial phase. The manufacturing methods proposed by the contractor in order to fulfil the technical requirements will be also addressed. (C) 2015 Elsevier B.V. All rights reserved.

Efficient development of a complex system such as a fusion component needs a stringent integration of standard and new constraints. For example, compared to the previous fusion experimental devices, remote handling (RH) and safety requirements are in ITER key parameters which must be integrated since the earliest design. For optimizing such integration studies, CEA, IRFM decided in 2010 to implement the use of virtual reality (VR) tools during the life cycle (from design to operation) of a fusion component. This paper describes a first feedback of such use for fusion engineering purposes.After a short overview of the CEA, IRFM VR platform capabilities, three main uses will be described: design review, simulation of remote handling and hands-on operations, with mat in the loop.The Design review mode was intensively used within the framework of a fruitful collaboration with ITER design Integration Team. This mode, fully compatible with CAD software, enables scale one data visualization with stereoscopic rendering. It improves the efficiency in detecting inconsistencies inside models and machine sub-system design optimization needs. Several accessibility cases of major Safety Important Components (SIC-1) were studied giving important requirements to the design at an early stage.CEA, IRFM, in close collaboration with expertise of CEA, LIST for VR simulation software, applies VR technologies for designing RH maintenance scenario for ITER Test Blanket System (TBS) and Ion cyclotron Resonance Heating (ICRH) Port Plugs. RH compatibility studies using VR pointed out major design drivers while helping to propose credible solution. VR platform is intensively used in the design of WEST (Tungsten (W) Environment Steady-state Tokamak) components and assembly studies, providing important information about the feasibility of assembly processes, optimization of physical mock-ups and ergonomic posture and gestures of operator.Finally, new perspectives, as the integration of safety constraints (dose calculation) will be described, demonstrating the powerful of VR tools at different stages of the component lifecycle. (C) 2015 Elsevier B.V. All rights reserved.

The Tore Supra tokamak is being transformed in an X-point divertor fusion device in the frame of the WEST project, launched in support to the Iter tungsten divertor strategy. The WEST Baffle aims to evacuate particles by creating a toroidal pumping throat. It also protects the lower divertor cooling pipes and the passive stabilization plate from heat fluxes. The Baffle is made of actively cooled plasma facing components (PFCs), with underneath a passive stabilization plate and a support beam. The Baffle design is presented in a first part. In a second time the different ANSYS simulations are described: electromagnetic, thermal and mechanical results are presented. The design choices are justified for the different results obtained thanks to the simulation. In a final part, some aspects of the PFC manufacturing are presented. (C) 2015 Elsevier B.V. All rights reserved.

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10-20 MW/m(2) range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program. WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER. (c) 2014 Elsevier B.V. All rights reserved.

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST project, launched in support to the ITER tungsten divertor strategy. The installation of coils inside the vacuum vessel led to the design of a divertor supporting platform able to meet the project requirements and the associated electromagnetic loads. This paper illustrates the design, the method and the results of the thermomechanical elastic stress analyses performed in 2012. The validation of the integrity of the structure is based on the compliance with RCCMR design criteria (even though these Design and Construction rules for Mechanical Components of nuclear installations are not required for such experimental fusion device). Several 3D analyses are performed with the ANSYS code. The major one is a global analysis of half structure which determinates the stresses in the main part of the components. It gives an idea of the areas which needs local analyses. It also provides the interface loads for junction studies or simplified local model. (C) 2014 Elsevier B.V. All rights reserved.

This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W - for tungsten - Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule. (C) 2014 Elsevier B.V. All rights reserved.

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.