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The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF. [ABSTRACT FROM AUTHOR]

• Since no existing RCC-MRx post-processing tool exists off the shelves, a specific methodology for pipe stress analysis is being developed by ITER, CEA and ASSYSTEM. • CEA and ASSYSTEM are jointly developing a specific methodology for pipe stress analysis in relation to the IO TBM section. • The Post-Processing tools for Pipe Stress Analysis Methodology are being developed with the intention of being open-source and accessible to all project participants. • The methodology described in this article was successfully applied to Pipe Forest Test Blanket Modules pipes routing. The ITER Test Blanket Modules (TBMs) are installed and operated inside the vacuum vessel (VV) at equatorial ports located within port plugs (PP). The Pipe Forest (PF) is a network of pipes connecting the Ancillary Equipment Unit (AEU) to the TBMs. Depending of the TBM, the Pipe Forest includes various piping system such as Water Coolant (350°, 188 bars), Helium Coolant (500°, 100 bars) Lithium Lead and Tritium Extraction System. During operating states, the Pipe Forest main function is to withstand loads in normal, incidental and accidental situations according to RCC-MRx code criteria (including design rules for mechanical components for fusion installations). It includes severe thermomechanical loadings due to the thermal expansion added to seismic and vertical displacements events. In addition, the PF layout must satisfy constructability in constraint environments and maximize the space available for man access, which constitutes a real challenge for design. Since no existing RCC-MRx post-processing tool exists off the shelves, a specific methodology for pipe stress analysis is being developed by ITER, CEA and ASSYSTEM. This methodology starts with a post-processing calculation tool of stresses and reaction forces for looping back into the structural calculation. Moreover, a specific extraction module from 3D model to processing tool allows the conversion of CAD data to calculation entry files. Finally, a feedback loop is also re-integrated into the core CAD model through an import module, enabling the tracking of displacements and evaluate the impact of incorporating pipe movements into the analysis. This paper describes the pipe stress analysis methodology and its application for Pipe Forest system design life cycle. It covers the historical evolution of the routing including introduction of expansion loops. More recently, it incorporates the integrated calculation of the entire port cell, spanning from the back of the TBM sets, up to the connection pipes of the vertical shaft. [ABSTRACT FROM AUTHOR]

Removal/reinstallation of all Test Blanket System (TBS) equipment present in the Port Cell is required during the ITER Long Term Shutdown. TBS shall be designed so that occupational radiation exposure can be As Low As Reasonably Achievable (ALARA) over the life of the plant to follow the ITER Policy. The expected level of radiation in this area still allows performing maintenance tasks hands-on. However, the cumulated dose could be significant for operators. Classical Dose Reduction Measure (DRM) is to deploy Remote Handling systems. Consequently, use of robotized equipment, remotely operated means or collaborative robotics, have been investigated. Taking advantage of new digital technologies such as digital assistances, is expected to help operators during complex remote operations under limited vision conditions. Experiments were performed on a set of three TBS maintenance representative tasks: remote visual inspection of DN80 pipe, dye penetrant testing operation on pipe and dexterity test. A panel of remote handling equipment operators of different skill level was selected and involved onto these tasks. The results prove without ambiguity that for all operators the quality of the task execution is significantly improved when using digital assistances [ABSTRACT FROM AUTHOR]

• To prevent from contamination, the use of Air Fed Suits shall be considered for maintenance operations. • The use of eXtended Reality is an asset all along the design process from feasibility to safe accessibility, maintainability and optimized design & methods. • To study the feasibility of the Test Blanket Modules replacement operations in Port Cells, a human-in-the-loop design method by means of Virtual and Augmented Reality is used at CEA IRFM. • This approach affords defining methodology & criteria for qualification of operation in AFS and validating by simulation the processes & interventions. • The IRFM XR Lab is tuned to perform integration and operation configurations studies by means of VR and AR simulations for TBM Port Cell. • Virtual and Augmented Reality are extensively used taking into account a large set of constraints (ALARA, human factors, ergonomics, accessibility...). To study the feasibility of the Test Blanket Modules (TBM) replacement operations in ITER Port Cells, CEA develops a mixed (digital & physical) validation by simulations approach with immersive technology to verify hands-on operations. Main validation criteria concern operator accessibility, posture, detailed procedure sequence and time spent at workstations. Such complex replacement operations will assume series of hands on tasks for both normal and abnormal situations, sometimes with protection such as Air-Fed Suit (AFS). After initial validation of key operation feasibility on digital mock-up, the goal is to optimize the process with minimization of occupational radiation exposure under the As Low As Reasonably Achievable (ALARA) requirement. Virtual reality tools provide relevant means to develop a credible replacement scenario in a context of complex configurations. The process includes features such as ergonomics assessment tool coupled to multi operators' collaboration , tangible tools and integration of advanced real time physics engine. Lessons learnt from the field of existing CEA facilities enable identifying key parameters related to workstation Humans factors. This first set of observations would help to define methodology and criteria for qualification of operation in AFS. The Digital simulation validation is deployed with a nuclear operator avatar in AFS to demonstrate Nuclear Maintenance Operation feasibility considering study of arduousness, accessibility and visibility thanks to real time motion capture. The recording of the whole simulation allows tracing and replaying for post-analysis data including body angulations, clash detection and operator field of view. The integration of this approach at an early phase of the design process enables validating technical options for replacement scenario. This leads to promising and beneficial results concerning design and integration of complex components in constraint working environments. This paper presents a review of the CEA studies regarding the extended virtual reality capabilities and results to support design TBM integration and replacement scenario within a nuclear environment framework. [ABSTRACT FROM AUTHOR]

One function of the ITER remote maintenance system is the transportation of in-vessel components and remote handling systems to and from the vacuum vessel and docking stations in the Hot Cell via dedicated galleries and lift. The cask and plug remote handling system (CPRHS) has been adopted as the solution to provide this nuclear confinement and transportation. This paper discusses the development of the conceptual design to-date and presents the processes being implemented to effectively control the subsequent CPRHS development. The CPRHS is a complex suite of systems with a significant number of interfaces with other ITER systems. Significant effort is being made to ensure that the system requirements are comprehensively defined and carefully managed and a feasible solution is developed – including planned and rescue operations. With the CPRHS performing a critical confinement function appropriate processes are being put in place to control the system development of the CPRHS. The expectation is that the combination of these factors will have a significant impact on the successful implementation of the CPRHS. [ABSTRACT FROM AUTHOR]

Abstract: The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components. Periodically or on request, the IVVS probes will be deployed into the Vacuum Vessel from their storage positions (still within the ITER primary confinement) in order to perform both viewing and metrology on plasma facing components (blanket, divertor, heating/diagnostic plugs, test blanket modules) and, more generically, to provide information on the status of the in-vessel components. In 2011, the IO proposed to simplify and strengthen the six IVVS port extensions situated at the divertor level. Among other important consequences, such as the relocation of the Glow Discharge Cleaning (GDC) electrodes at other levels of the machine, this major design change implied the need for a substantial redesign of the IVVS plug, which took part to an on-going effort to bring the integrated IVVS concept – including the scanning probe and its deployment system – to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the various design and R&D activities in progress: plug design integration, probe concept validation under environmental conditions, development of a metrology strategy, the whole supported by a nuclear analysis. [Copyright &y& Elsevier]

Abstract: The availability of the ITER machine to perform its scientific program is strongly dependent on the performance of the different Remote Handling (RH) systems constituting the ITER Remote Maintenance System (IRMS). The lifecycle of the IRMS will largely exceed 40 years from initial concept design and proof testing through to machine decommissioning. Such a long lifecycle requires that a rigorous approach is put in place to guarantee the technical capabilities of the highly innovative IRMS, its efficiency and its availability. For this purpose, an IRMS System Engineering and IRMS lifecycle management approach has been adopted by ITER. The approach aims at ensuring the IRMS full operability and availability at an acceptable cost of ownership over the full ITER machine assembly and operations period. The IRMS lifecycle management method described in this paper covers such subjects as specific requirements for IRMS design reviews, monitoring during manufacture, factory and site acceptance testing, integrated commissioning, decontamination, maintenance and re-qualification strategies, requirements for Integrated Logistical Support during operations. The updating and implementation of the IRMS lifecycle strategy and this procedure will be managed and monitored by the Remote Handling Integrated Product Team (RH-IPT). Although developed for the IRMS, the basic principles and procedures of lifecycle management could be applied to other ITER plant systems whose reliability and availability will be essential for the continued operation of the ITER machine. [Copyright &y& Elsevier]

Abstract: In the context of ITER, CEA/IRFM has participated to the design and integration of several components in the Equatorial Port plug region. Particularly, in the framework of the grant F4E-2008-GRT-09-PNS-TBM, CEA/IRFM has contributed to the test blanket module system (TBS) design and robot access feasibility study in the Port Cell. Simulations of the maintenance procedure were studied and fully integrated to the design process, enabling to provide space reservation for human and robotic access. For this mean, CEA/IRFM has used a CEA LIST Virtual Reality simulation software directly integrated to the Solidworks CAD software. The feasibility to connect/dis-connect the pipes in front of the Bioshield by a set of potential standard industrial arms was demonstrated. Aiming to give more realism to maintenance scenario and CAD models, CEA IRFM has decided to build a Virtual Reality platform in the institute, integrated to the design office. With the expertise of CEA LIST, this platform aims to provide the nearest possible links between design and remote handling needs. This paper presents the outcome of the robot access study and discusses about the Virtual Reality tools that are being developed for these applications. [Copyright &y& Elsevier]

Abstract: The NB ITER Remote Maintenance System (NB IRMS) provides the means for the remote maintenance within the NB Cell by removal and replacement of the plant equipment. The NB IRMS will be installed and removed with the assistance of human workers during the preparation, and post-operation phase. During the maintenance operation after opening the Passive Magnetic Shield (PMS) and vessels, the maintenance activity and recovery from failure should be conducted remotely. This paper describes the concept design of the NB IRMS operating inside the NB cell for maintenance of the plant equipment such as NB components, and Upper Port Plugs (UPP). The main tasks of the IRMS, the description of the sub-systems and their specification, and deployment/operation principles are presented. The transportation concept of the NB IRMS to the hot cell facility for storage and maintenance is presented, which is to avoid unnecessary exposure on the equipment inside the NB cell during the machine operation. [Copyright &y& Elsevier]

Abstract: 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 10kg 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. [Copyright &y& Elsevier]

• The ITER TBM Program plans to test DEMO-relevant tritium breeding blanket systems. • The selection of the 4 TBSs for Initial Configuration is expected in a near future. • Each TBS is formed by various sub-systems with different functions and technologies. • The selection of the four InCo TBSs is based on the existing TBS conceptual designs. • The infrastructures hosting the TBSs has reached the preliminary design stage. The ITER Test Blanket Module (TBM) Program has significantly evolved since 2018. The number of equatorial ports allocated for operating the Test Blanket Systems (TBSs) has been reduced from three to two. As consequence, four TBSs can be simultaneously installed and operated, versus six previously. Since the dedicated space in the various rooms of the Tokamak Complex has been kept unchanged, the existing space constraints have been substantially relaxed. The paper addresses the possible selection of the four TBSs for the initial configuration and design of the involved TBMs and the main on-going R&D carried out by the ITER Members (IMs) in support of the TBS designs. It describes also the main progress of the design of the infrastructures needed for hosting the four TBSs (e.g., port plugs, port cell common components, common maintenance tools and equipment). The design of the TBS Connection Pipes (CPs) System has reached its final phase since, being captive, it needs to be installed before the First Plasma. [ABSTRACT FROM AUTHOR]