Achievement of an economic fusion reactor imposes a high level of unprecedented requirements for neutral beam systems; the first one is the nuclear safety constraints which impose a prerequisite on the ease of access of all injector components to provide a preventive (and curative) maintenance by remote handling while maintaining reactor operation. In addition, the challenge is to develop high-power injectors with very high wall-plug efficiency (above 60%) that are able to operate in stable conditions over several months. There is a significant gap to bridge with respect to the present neutral beam systems, which are characterized by a low efficiency and by complex and long maintenance operations. Evidence that this injector concept does not offer adaptations to cope with the reactor requirements makes it clear that a new concept has to be addressed. An injector concept with modular sources at ground voltage is proposed. The concept makes remote maintenance of the injector components easier, where each source module can be replaced by a new one without breaking the vacuum and affecting injector conditioning. With the grounded and modular ion source, photoneutralization associated with energy recovery appears to be the best route capable of attaining the reactor requirements. This concept of a maintainable neutral beam system would provide a high heating power with a wall-plug efficiency above 70% and unprecedented features such as the capacity of producing temporal and spatial modulation of the beam power for a better control of the plasma stability. Up to now, photoneutralization feasibility studies already carried out on reduced-scale prototypes have not highlighted any showstoppers. Continuation of research and development in the years to come can pave the way towards the achievement of a first full-scale high-power cavity in the 1 MW range, and the realization of a multi-amperes (∼10 A) thin blade-like D − beam.

Currently on fusion devices, diagnostics are mainly aiming at plasma analysis and control. However, operational and programmatic needs have appeared for regular in-vessel components monitoring during plasma campaign. Light robotics systems could meet this requirement and may be a way as well to replace human interventions to fix damaged in vessel components. To minimize the impact on machine operation, the robotic system has to be mini-invasive and compatible with operating conditions (vacuum, temperature.). To fulfill this goal, CEA has developed a multipurpose carrier able to be operated inside WEST vessel between plasma pulses. A prototype of this robot, called Articulated Inspection Arm (AIA), was tested in 2008 in Tore Supra vacuum vessel. A major upgrade was performed in 2014-2015 with the aim of converting this prototype into a reliable tool in support to WEST operation. During the WEST components manufacturing and installation (2014-2016), the robot was integrated and tested in the EAST Tokamak. Since 2017, the AIA has been regularly used during the WEST plasma campaigns. Movies provided by the embedded camera allow to assess the evolution of Plasma Facing Components surface state and the effects of plasma loads, runaways and disruptions. The robot operation was also very helpful to assess the needs for maintenance, to assist mechanical assembly without man entry and to perform diagnostics calibration under relevant conditions.

The new tokamak generation will be characterized by the necessity of full remote maintenance for most of the critical components. Because no human intervention can be envisaged, the remote systems will have to prove high reliability and rescue capabilities. The availability of the fusion facility will have to be maximized, as a consequence, the efficiency of the maintenance equipment will become a key factor. To fulfill these constraints, the remote maintenance principles will need to be simplified as much as possible and be taken into account at the early stages of the tokamak design. This is an iterative process between Tokamak and Maintenance designs in order to reach the best possible tradeoff. Such an exercise of this iterative work was done on the China Fusion Engineering Test Reactor (CFETR) preliminary design. Main targets were established in the view of optimization of the remote maintenance, then an assessment of the first maintenance scenario, envisaged for the in-vessel components, was done and alternative solutions were proposed. This process was repeated during meetings with the tokamak subsystem owners. This paper will present the results of a first phase of optimization that can represent the current state of the remote maintenance design for CFETR. The work has provided guidelines for main tokamak subsystems such as the magnet configuration, the blanket and divertor modularity, the neutron shielding and cryostat arrangement, and the transfer and hot cell configuration. It also produced some innovative solutions for the overall maintenance scheme like the implementation of a hot cell at the top of the tokamak hall avoiding cask transfers. On this basis a CFETR maintenance scenario is proposed.

The EU-DEMO remote maintenance strategy must be relevant for a range of in-vessel component design options. The remote maintenance project must provide an understanding of the limits of the strategy and technologies so as to inform the developing plant design of the maintenance constraints. A comprehensive set of maintenance requirements has been produced, in conjunction with the plant designers, against which design options can be assessed.The proposed maintenance solutions are based around a strategy that deploys casks above each of the vertical ports to exchange the blanket segments and at each of the divertor ports to exchange the divertor cassettes. The casks deploy remote handling equipment to open and close the vacuum vessel, remove and re-install pipework, and replace the in-vessel components.A technical design risk assessment has shown that the largest risks are common to all of the proposed solutions and that they are associated with two key issues, first; the ability to handle the large blanket and divertor components to the required positional accuracy with limited viewing and position feedback, and second; to perform rapid and reliable pipe connections, close to the blankets, with demonstrated quality that meets the safety requirements. (C) 2016 Published by Elsevier B.V.

Fusion Tokamaks are complex machines which require special conditions for their operation, in particular, high vacuum inside the vessel and high temperature of the vessel walls. During plasma phases, the first wall components are highly stressed and a control is necessary in case of doubt about their condition. To be able to make safely such an inspection in a short period of time is a great advantage. The Articulated Inspection Arm (AIA) developed by the CEA provides the capability for fast inspections of the first wall overall surface keeping the vacuum and temperature conditions of the vessel. The robot prototype was validated in Tore Supra in 2008. In the frame of a joint laboratory, CEA/IRFM and ASIPP have decided to upgrade the existing AIA prototype to use it routinely in the EAST and WEST tokamaks. The robot has followed an important upgrade program in 2013 and 2014. The document presents the various upgrades made on the mechanics, the sensors, the electronics, the control station and the integration adaptation for the operation on EAST. From the AIA experience, thoughts for future inspection robots are given. (C) 2015 Elsevier B.V. All rights reserved.

An articulated inspection arm (AIA) has been developed by CEA for visual inspection between pulses inside the Tore Supra tokamak vacuum vessel without breaking temperature and vacuum conditions. The eight meters length robot is composed of a shuttle and six articulated segments with a video camera at its end. A demonstration prototype has been achieved in 2008 at Tore Supra (Gargiulo, 2007; Houry, 2008; Perrot, 2003).A project to upgrade the AIA into a fully operational robot has been undertaken by IRFM and ASIPP in an Associated Laboratory. It will be in operation first in the EAST machine and afterwards in Tore Supra in its WEST (W/Tungsten Environment in Steady-state Tokamak) configuration where it is of paramount importance to survey possible degradation of W component surface.The control system of the robot has been extensively upgraded. The effort has been focused on three areas: (1) improvement of the arm position accuracy, (2) increase of the operational robustness, (3) use of a powerful graphical user interface including simulation of trajectories and robot deployment capabilities in a 3D viewer environment.The aim of this paper is to detail the architecture of the AIA control system. (C) 2015 Elsevier B.V. All rights reserved.

EAST Articulated Inspection Arm (AIA) robot is being mutually developed by ASIPP and CEA-IRFM for remote handling maintenance. It will permit remote visual inspection and to pick up small fragments inside the EAST tokamak vacuum vessel during experiments. Considering storage and support for EAST AIA, a sealed cask system has been designed and manufactured, which can be connected to EAST device through a ϕ250 mm connection port with two flashboard valves. The system consists of a 10 m long vacuum vessel with a linear guide rail for storage, guiding and conditioning, two mobile wagons for support and some auxiliary systems for keeping suitable work conditions and measurement. Besides, a stainless steel shuttle has been developed to support AIA robot and assemble with the linear guide. It can push the robot into tokamak vessel and back to the storage cask with a gear-rack driving mechanism. This paper mainly presents the overall description of the system design and some obtained implementation progress.

The aim of the JET ITER-like Wall Project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.

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.

As the central part of a programme of enhancements in support of ITER, the Joint European Torus (JET) is being equipped with an all-metal wall. This enhancement programme requires the removal and installation of 6927 tile carriers and tiles, as well as the removal and installation of embedded diagnostics and antennas. The scale of this operation and the necessity to maximise operational availability of the facility added a requirement for high productivity in the remote activities to the existing exigencies of precision, reliability, cleanliness and operational security. This high productivity requirement has been incorporated into the design of the components and associated installation tooling, the design of the installation equipment, the development of installation procedures including the use of a mock-up for optimisation and training.Consideration of the remote handling installation process is vital during the design of the in vessel components. A number of features to meet the need of the high productivity while maintaining the function requirements have been incorporated into the metal wall components and associated tooling including kinematic design with guidance appropriate for remote operation. The component and tools are designed to guide the attachment of the installation tool, the installation path, and the interlocking with adjacent components without contact between the fragile castellated beryllium of the adjacent tiles. Other incorporated ergonomic features are discussed.At JET, the remote maintenance is conducted using end effectors, normally hi-lateral force feed back manipulator, mounted on driven, articulated booms. Prior to the current shutdown one long boom was used to conduct the installation and collect and deliver components to the "short" boom [3] which was linked to the tile carrier transfer facility. This led to loss of efficiency during these movements. The adoption of a new remote handling philosophy using 'point of installation' delivery of components via an additional long boom and a sophisticated logistics system based on 'task modules' is described and operational efficiencies detailed. The enabling programs and software behind this new approach needed significant development. Systems such as a task module manager database, image visualisation, virtual reality simulation, operational document system and teach files. which have significantly evolved since their initial inception in 2002, will be elaborated. Crown Copyright (C) 2011 Published by Elsevier B.V. All rights reserved.

This paper reports the successful installation of the JET ITER-like wall and the realization of its technical objectives. It also presents an overview of the planned experimental programme which has been optimized to exploit the new wall and other JET enhancements in 2011/12.

A precise geometric method is used to calculate the power deposition on the future JET ITER-Like Wall beryllium tiles with particular emphasis on the internal edge loads. If over-heated surfaces are identified, these can be modified before the machining or failing that actively monitored during operations. This paper presents the methodology applied to the assessment of the main chamber beryllium limiters. The detailed analysis of one limiter is described. The conclusion of this study is that operation will not be limited by edges exposed to plasma convective loads.

This paper presents an overview of the status and relevant technical issues for the ITER-like Wall Project with emphasis on progress since the 11th International Workshop on Plasma-Facing Materials and Components for Fusion Applications.

The ITER-like wall project in JET aims at an optimal use of the unique features of JET, such as beryllium and tritium compatibility, to explore operation within the limits of the ITER wall materials. A full replacement of the presently carbon-based first wall will result in the surfaces facing the plasma being mainly beryllium in the main chamber and tungsten in the divertor. At the same time, the JET auxiliary heating power will be upgraded allowing access to ITER-relevant energy loss densities in disruptions and edge localized modes. In this way, the JET wall will go from being almost indestructible, to making the material-driven operational constraints predicted for ITER a more immediate reality for JET. This paper describes the methodology being used to define and apply these limits while optimizing experimental flexibility.

ITER-Like Wall Project has been launched at the JET tokamak in order to study a tokamak operation with beryllium components on the main chamber wall and tungsten in the divertor. To perform this first comprehensive test of both materials in a thermonuclear fusion environment, a broad program has been undertaken to develop plasma-facing components and assess their performance under high power loads. The paper provides a concise report on scientific and technical issues in the development of a beryllium first wall at JET.

Currently, the primary ITER materials choice is a full beryllium main wall with carbon fibre composite at the divertor strike points and tungsten on the upper vertical targets and dome. The full tungsten divertor option is a possibility for the subsequent D–T phase. Neither of the ITER material combinations of first wall and divertor materials has ever been tested in a tokamak. To collect operational experience at JET with ITER relevant material combination (Be, C and W) would reduce uncertainties and focus the preparation for ITER operations. Therefore, the ITER-like wall project has been launched to install in JET a tungsten divertor and a beryllium main wall. This paper describes the R&D activities carried out for the project to develop an inertially cooled bulk tungsten divertor tile, to fully characterise tungsten coating technologies for CFC divertor tiles and to develop erosion markers for use as diagnostics on beryllium tiles.