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• Methodology to perform RE activities during ITER construction is presented. It covers the Tokamak Machine and Buildings. • Reverse Engineering implementation workflow is described. • RE activities are implemented and managed in ITER by the As-Built Data management Working Group (ADM-WG). Group organization is described. • RE analysis is a valuable tool for assembly process optimization, interfaces customization and alignment target optimization. • RE methodology is described. The application of RE techniques to the assessment of Vacuum Vessel and Cryostat Base construction is presented. The ITER machine consists of a large number of highly integrated and complex systems, with critical functional positional requirements (e.g. accurate positioning of magnets to minimize error fields and location of plasma facing components with respect to magnetic axis) and reduced design clearances to maximize Tokamak performances and optimize sizing. Deviations from specified part tolerances and product assembly processes/accuracies could have a critical impact on assembly feasibility, schedule and achievement of final performances, compromising plasma operation. The assessment of potential deviations on as-manufactured and as-assembled systems, together with early mitigation of non-compliance with Tokamak dimensional requirements, are critical activities during the ongoing manufacturing and Tokamak construction phases. This paper describes the methodology implemented within ITER for the assessment of as-built integrated systems, based on Reverse Engineering (RE) techniques. It includes the identification of construction needs (at system and assembly levels), the acquisition of metrology data based on defined needs, the post-processing and alignment of data and the reconstruction and management of 3D as-built configuration models. RE techniques are a basic tool for the management of dimensional risk and opportunity for mitigation or recovery during ITER construction phase. ITER is a Nuclear Facility INB-174. This paper describes a Protection Important Activity (PIA) for safety. [ABSTRACT FROM AUTHOR]

Highlights • Development of a simple and standardized handling process for ITER Port Plugs using passive guiding means. • Pre-selection of materials and low-friction treatment, compatible with ITER in-Vessel environment, based on Tribology tests. • Demonstration of ITER Port Plug Handling process mechanical soundness by successful trials on a physical downscaled mock-up. • Optimization of ITER Port Plug gap configuration based on 3D Tolerance analysis (3DCS). Abstract The Port Plug handling system aims to insert and remove the ITER Port Plugs installed at the Equatorial and Upper levels of the ITER Tokamak. As the Port Plugs activation and contamination levels prevent hands-on access, their safe transfer is ensured by the Cask and Plug Remote Handling System (CPRHS) between the buildings. The basic principle of the CPRHS is to allow the transfer of the Plug in and out the vessel using passive guiding means permanently attached to the Plug and the Vessel. This handling process has been reproduced on a scaled-down physical mock-up which has partially demonstrated its reliability from a mechanical point of view. In parallel, a campaign of tribology testing has been performed to identify suitable material compatible with the In-Vessel environmental conditions and with which to optimize the coefficient of friction. This paper describes the technical activities performed during handling process development (integration, tolerances, mechanical and neutronic analyses) and the results of performed tests. [ABSTRACT FROM AUTHOR]

The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed at mid-2014. The paper gives an overview of the final configuration of the ITER nuclear buildings and highlights the large progress made on the final integration of the plant systems in the Tokamak complex. The revised design of the Tokamak machine supporting system is also described. [ABSTRACT FROM AUTHOR]

The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on the Tokamak Dimensional Variation Model is also reported. ITER is a Nuclear Facility INB-174 (Installation Nuclear de Base no. 174). This paper describes a Protection Important Activity (PIA) for safety. [ABSTRACT FROM AUTHOR]

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill this requirement, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) is composed of several test stands that can be used to test the port plugs whereas at the end of manufacturing (in a non-nuclear environment), or after refurbishment in the ITER hot cell facility. The PPTF provides the possibility to perform environmental (leak tightness, vacuum and thermo-hydraulic performances) and functional tests (radio frequency acceptance tests, behavior of the plugs’ steering mechanism and calibration of diagnostics) on upper and equatorial port plugs. The final design of the port plug test facility is described. The configuration of the standalone test stands and the integration in the hot cell facility are presented. [ABSTRACT FROM AUTHOR]

Abstract: The optimization of the manufacturing/assembly tolerances and processes in ITER Experimental Nuclear Fusion Device is one of the key tasks to optimize the fabrication cost, to prevent problems during assembly and to ensure that the critical homogeneity of the magnetic field and the positioning requirements of the plasma facing components can be achieved. This task is further complicated by the strong interplay among the various Tokamak systems, as for instance in the inner region of the machine where the clearances between Central Solenoid, Toroidal Field Coils, Thermal Shield, Vacuum Vessel and In-Vessel components have been minimized for their large influence on the magnetic flux and the overall machine cost. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. The dimensional variation model is representative of Tokamak functional tolerances and processes, predicting accurate values for the amount of variation on critical areas. In addition, dimensional simulations help to determine the key tolerances that contribute to a particular variation. This paper describes the current status of the Tokamak dimensional variation studies and its management plan, highlighting the status of compliance of allocated tolerances with input requirements. Management of risk issues and corrective actions are also described. [Copyright &y& Elsevier]

Abstract: The ITER site consists of almost 30 buildings to service the Tokamak machine which is located in the centre of the Tokamak Complex facility with the Tokamak-, Diagnostic- and Tritium building. The design of a large part of the ITER plant systems will be executed by the ITER Domestic Agencies or their industrial suppliers under functional specifications provided by the ITER Organization. At the same time, the detailed design of the building is carried out by the European Domestic Agency ‘Fusion for Energy’ (F4E). In order to allow an efficient identification of the ITER configuration as well as to manage the concurrent engineering activities and to simplify the identification and assessment of changes, the design of each ITER plant systems is described in the so-called Configuration Management Models (CMM). These are light CATIA® 3D models that define the required space envelope and the physical interfaces in-between the systems and the buildings. The paper describes the procedure adopted for the control of the baseline configuration of the Tokamak Complex facility and Auxiliary Buildings with their associated plant systems and illustrates the current status as well as recent developments in the different systems. [Copyright &y& Elsevier]

Abstract: In the field of the ITER port plug engineering and integration task, CEA has contributed to define proposals concerning the port plugs vacuum sealing interface with the vessel flange and the equatorial plug handling. The 2001 baseline vacuum flange sealing consisted of TIG welding of a 316L strip plate on to U shapes. This arrangement presented some issues like welding access, implementation of tools, lip consumption, complex local leak test, continuous leak checking. Therefore, an alternate sealing solution based on the use of metallic gaskets is proposed. The different technical aspects are discussed to explain how this design can simplify the maintenance and deal with safety and vacuum requirements. The design of the mechanical attachment and vacuum sealing of the plug has constantly evolved, but the associated remote handling equipment was not systematically reviewed. An update of the cask and maintenance procedure was studied in order to design it in accordance with the last generic plug flange design. This includes a concept of a gripping system that uses the plug flange bolting area and, to help the remote handling process, a cantilever assisting system is suggested to increase the reliability of the transfer operation between vacuum vessel and cask. [Copyright &y& Elsevier]

Abstract: Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in-vessel inspection operations without loss of conditioning will be mandatory. In this context, an Articulated Inspection Arm (AIA) is currently developed by the CEA within the European work programme framework, which aims at demonstrating the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10kg). It is composed of 5 segments with 8 degrees of freedom and a total range of 8m. The first in situ tests will take place by the end of 2007 on the Tore Supra Tokamak at Cadarache (France). They will validate concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for various applications. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: [•] The first embedded process is the viewing system. It is already manufactured and will allow close visual inspection of the complex Plasma Facing Components (PFC) (limiters, neutralisers, RF antenna, diagnostic windows, etc.). [•] In situ localisation of water leakage based on a helium sniffing system is also being studied to improve and facilitate maintenance operations. [•] Finally a laser ablation system for PFC detritiation, developed in CEA laboratories, is being fitted to be implemented on the robot for future operation in Tore Supra. This paper deals with the integration of the robot into Tore Supra and the progress in the development of the processes listed above. It also describes the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions. [Copyright &y& Elsevier]