Structural analysis of ITER sub-assembly tools

The ITER Tokamak assembly tools are purpose-built assembly tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. The sector sub-assembly tools descried in this paper are main assembly tools to assemble vacuum vessel, thermal shield and toroidal filed coils into a complete 40° sector. The 40° sector sub-assembly tools are composed of sector sub-assembly tool, including radial beam, vacuum vessel supports and mid-plane brace tools. These tools shall have sufficient strength to transport and handle heavy weight of the ITER Tokamak machine reached several hundred tons. Therefore these tools should be designed and analyzed to confirm both the strength and structural stability even in the case of conservative assumptions. To verify structural stabilities of the sector sub-assembly tools in terms of strength and deflection, ANSYS code was used for linear static analysis. The results of the analysis show that these tools are designed with sufficient strength and stiffness. The conceptual designs of these tools are briefly described in this paper also.     

Magnetic compatibility of standard components for electrical installations: Computation of the background field and consequences on the design of the electrical distribution boards and control boards for the ITER Tokamak building

Inside the proposed Tokamak building, the ITER poloidal field magnet system would produce a stray magnetic field up to 70 mT. This is a very unusual environmental condition for electrical installation equipment and limited information is available on the magnetic compatibility of standard components for electrical distribution boards and control boards. Because this information is a necessary input for the design of the electrical installation inside the proposed ITER Tokamak building specific investigations have been carried out by the ITER European Participant Team. The paper reports on the computation of the background magnetic field map inside the ITER Tokamak building and the consequences on the design of the electrical installations of this building. The effects of the steel inside the building structure and the feasibility of magnetic shields for electrical distribution boards and control boards are also reported in the paper. The results of the test campaigns on the magnetic field compatibility of standard components for electrical distribution boards and control boards are reported in companion papers published in these proceedings.     

Progress in the design of ITER front-end vacuum instrumentation and controls

The ITER vacuum system will be one of the largest and most complex vacuum systems ever to be built. Extensive instrumentation and controls are being developed to satisfy the stringent vacuum processes necessary for the successful and safe operation of the ITER Tokamak. The complexity and deep integration of the vacuum systems within the ITER machine presents a challenge to implement all of the controls necessary for reliable operation. Several thousand valves and sensors have to be implemented within the harsh environmental conditions of the Tokamak vicinity, and require engineering of instrumentation and controls with remote electronics solutions.In this paper the status of the design of field end vacuum controls and instrumentation for the ITER vacuum systems is described. Details of the progress on selection of sensors and actuator technologies are given herein and solutions for remote device operation, including those for cryogenic devices, are described together with necessary local shielding.     

ITER Hot Cell optimisation

The Hot Cell has the pivotal role in supporting on-going maintenance of the ITER machine. The experimental nature of the ITER Tokamak dictates that Hot Cell tasks will be complex in nature and will evolve over the lifetime of the project. This paper presents the results of a study commissioned by the ITER Organisation and undertaken by Oxford Technologies Ltd and Comex Nucléaire to investigate optimisation of the Hot Cell baseline design to ensure it is able to fulfil these challenging requirements.The study reviews the rationale for the current baseline Hot Cell design, derives a set of design variables which have a significant influence on the Hot Cell (e.g. Divertor refurbishment strategy) and assesses the impact of these variables. The method adopted was to perform a quantitative analysis of the impact of each variable individually without consideration of any other factors. Following on from this a multi-dimensional analysis, performed by expert assessment, was undertaken to assess the relative benefits and limitations of all of these interacting variables. The output of this study are alternative Hot Cell layouts which are optimised for criteria considered critical to the ITER Organisation, i.e. maximising Tokamak availability and minimising Hot Cell cost and size. This paper discusses the overall study process, issues raised and results.     

Design modification and optimization of the ITER cooling water system

ITER (Latin for “the way”), the largest fusion experimental reactor in the world, is designed to demonstrate the technological feasibility of nuclear fusion energy conversion, at plant scale, from high temperature deuterium-tritium plasma using the Tokamak magnetic confinement arrangement.ITER will have a large vacuum vessel that hosts the plasma facing components. These components include the blanket and the divertor that will operate at temperatures, heat loads, and neutron flux higher than those reached in a nuclear fission power plant reactor.One of the main critical issues of the ITER reactor is the design of the cooling water system to transfer the heat generated in the plasma to the in-vessel components and the heat loads from the auxiliary systems to the environment.This paper describes the current ITER cooling water system and recent design modifications and optimizations.     

Conceptual design of the hot cell facility universal docking station at ITER

Between main shutdowns of the ITER machine, in-vessel components and Iter Remote Maintenance System (IRMS) are transferred between the Tokamak complex and the Hot Cell Facility using different types of sealed casks. Transfer Casks have different physical interfaces with the Vacuum Vessel, which need to be the same at the docking stations of the HCF. It means that in-vessel components and IRMS are cleaned in the same cells, which is in fact not convenient. Furthermore, logistic studies showed that the use rate of the cells is very inhomogeneous. In order to have dedicated cell for decontamination of Remote Handling tools, in order to increase the operability efficiency and to removes the hot cell docking operation from the critical path, the concept of a universal docking station has been investigated. Based on an existing design, the work was focused on a review of requirements, the re-design and the integration within the HCF layout. The universal docking station has been proposed and is now integrated in HCF design.     

Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.     

Development on systems configuration in ITER Tokamak Complex and Auxiliary Buildings

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.     

Study on the Flexible Support Legs for ITER Transfer Cask System

In fields of remote handing i.e. robot technology for fusion engineering reactor, such as ITER or the China fusion engineering test reactor, the flexible support legs are key components for their transfer cask system to adjust its position, joining to hot cell or tokamak ports for maintaining the fusion device. For ITER machine, each support leg should withstand maximum 50 tons load and adjust its height in 150 mm. Defect in original ITER design was presented. A new concept for the support legs was configured and its feasibility was proven. Detailed design and simulation was done for the new support leg with virtual prototype technology. Simulation results show that new support leg could not only meet all required function but also has merits of constant load during the tuning process with linear relation of control variable parameters, which is intended to be used for Tokamak reactors.     

Production and installation study of the ITER PF4 in-pit feeder

The PF4 in-pit feeder includes the In-Cryostat-Feeder (ICF) and Cryostat-Feed-Through (CFT), with busbars being their key components. The relative positions of the busbar terminal joints are measured by using a laser tracker and adjusted by positioning tools. The busbars are not fixed on the separate plate until the position errors meet the manufacture tolerances. The CFT has a 2.38° penetration angle relative to the ground and will be installed firstly. The position of the connection interface between the CFT and its lifting tool is analyzed, and to reduce the total deformation and keep the assembly precision of the joints the straight part of the CFT needs to be supported. The ICF has the most critical assembly operation space, as it must be installed on its temporary support in a temporary position. After the PF4 coil has been installed the ICF will be moved to its final position. To guarantee the 30 mm safety assembly clearance provided by ITER, the collision analysis of the ICF is performed, which demonstrates that the assembly procedures are feasible.     

The WEST project: Current status of the ITER-like tungsten divertor

The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units.The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.     

Magnetic compatibility of standard components for electrical installations: Tests on low voltage circuit breakers and contactors

The components of the electrical distribution system installed in the ITER Tokamak Building are subjected to a constant or slowly variable magnetic field, an environmental condition unusual for standard components used in low voltage electric distribution systems. Wide typologies of breakers, contactors and protection relays and a complete Low Voltage Distribution Board have been extensively tested to assess their behaviour when subjected to magnetic field. The results allowed to establish a magnetic field limit for their regular operation and to investigate upon overall functionality of a Low Voltage Distribution Board and upon the shielding effect of its iron structure.     

A programmatic framework for the Tokamak Physics Experiment (TPX)

Significant advances have been made in the confinement of reactor-grade plasmas, so that we are now preparing for experiments at the power breakeven level in the JET and TFTR experiments. In ITER we will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, we must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the U.S. domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX we can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. We can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current U.S. fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program.Abbreviations used TFTR Tokamak Fusion Test Reactor - JET Joint European Torus - ARIES Advanced Reactor Innovations Evaluation Study - SSTR Steady State Tokamak Reactor - PBXM Princeton Beta Experiment-Modified - DIII-D Doublet III—Dee - JT60-U Japanese Tokamak 60-Upgrade     

ITER blanket manifold system: Integration,assembly and maintenance

The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned.     

Electromagnetic analysis, structural integrity and progress on mechanical design of the ITER ferromagnetic insert

Ferromagnetic material is used to reduce the toroidal field ripple in JFT-2M [H. Kawashima, et al., Demonstration of ripple reduction by ferritic steel board insertion in JFT-2M, Nucl. Fusion, 41 (2001) 257-263] and JT-60U [H. Takenaga, the JT-60 Team, Overview of JT-60U results for development of steady-state advanced Tokamak scenario, Proceedings of the 21st IAEA Fusion Energy Conference, Chengdu, China, 2006]. In ITER, since the ferromagnetic material is inserted in the space between the double walls of ITER Vacuum Vessel (VV), it is called “ferromagnetic inserts”. Suitable material is selected to satisfy the design requirements of ITER. The proper location and amount of the ferromagnetic inserts are optimized with the goal of reduction of the toroidal field ripple. The ferromagnetic inserts are designed to minimize electromagnetic forces acting on them. The electromagnetic forces have been calculated with the latest disruption scenarios. Magnetization forces due to magnetic fields have also been calculated. Structural integrity has been validated by a structural analysis.     

ITER Cryostat—An overview and design progress

The ITER Cryostat is one of the most important and critical systems in the ITER project. It envelops the entire basic systems of the Tokamak and is a vacuum tight container. Cryostat provides vacuum environment for the thermal insulation to magnet system operating at 4.5k and thermal shield system operating at 80k. It is evacuated to a pressure of 10^?4 Pa prior to cool down of the magnets and thermal shields in order to limit heat transfer by gas conduction and convection to a level tolerable to the cryogenically cooled components. The cryostat is also designed to support all the loads like gravity, electromagnetic forces, seismic, etc. that derive from the Tokamak basic systems, and from the Cryostat itself, to the floor of the Tokamak pit through its support structures during the normal and off-normal operational regimes, and at specified accidental conditions. ITER Cryostat conceptual design was reviewed last year and detailed design is reviewed in June 2010. At present Cryostat design is in its final stage of completion and final design review is planned in October 2010 to finalize the procurement specifications. Procurement Arrangement will be signed this year with Indian Domestic Agency and procurement cycle will start for the fabrication of the ITER Cryostat. This paper discusses the updated Cryostat design and analysis with integration of the different penetrations required for the communication within the In-Cryostat system and its maintenance.     

ITER components cooling: Satisfying the distinct needs of systems and components

The ITER Tokamak requires multiple auxiliary systems to initiate, support, and monitor the fusion reaction. Heat produced by these systems, as well as the heat produced by the fusion reaction itself is collected by the ITER Cooling Water System (CWS) and rejected to the atmosphere. The CWS is composed of several systems designed for specific cooling roles. One of these systems is the Component Cooling Water System 2 (CCWS-2) whose function is to collect the heat from auxiliary client systems and components and transfer it to the Heat Rejection System. Clients are located throughout the site and have different requirements in terms of pressure, temperature, temperature variation, flow, metallurgy of wetted surfaces, and water quality. To satisfy these different requirements the CCWS-2 is divided into four separate loops, each of which has different operating parameters. For example, the CCWS-2A loop is designed to cool components with wetted surfaces of copper and primarily serves the radio-frequency heating systems, magnet power supplies, and neutral beam injector system components. This paper describes the evolution of the CCWS-2 system to match the needs of groups of compatible clients, and describes the development of the preliminary design of one of its loops, CCWS-2A, to meet individual client needs.     

Status of the ITER Ion Cyclotron H&CD system

The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40–55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R&D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper highlights progress and outstanding issues for the various system components.     

The design of the assembly tools for the ITER tokamak

The conceptual design of the purpose-built assembly tools required for ITER tokamak assembly is given. The ITER machine assembly is sub-divided into five major activities: lower cryostat, sector sub-assembly, sector assembly, ex-vessel, and in-vessel [1]. The core components, vacuum vessel (VV) and toroidal field coil (TFC), are assembled from nine 40° sub-assemblies, each comprising a 40° VV sector, two TFCs, and the associated VV thermal shield (VVTS). The lower cryostat activities must be completed prior to sector assembly in pit to prepare the foundations for the core components, and to locate the lower components to be trapped once the core components installation begins. In-vessel and ex-vessel activities follow completion of sector assembly. To perform these assembly activities requires both massive, purpose-built tools, and standard heavy handling and support tools. The tools have the capability of supporting and adjusting the largest of the ITER components; with maximum linear dimension 19 m and mass 1200 tonne, with a precision in the low mm range. Conceptual designs for these tools have been elaborated with the collaboration of the Korean Domestic Agency (KO DA). The structural analysis was performed as well using ANSYS code.     

The Alignment and Assembly for EAST Tokamak Device

EAST （HT-7U） is a large fusion experimental device. It is a full superconducting tokamak with 1 MA of plasma current, 1000 s of plasma duration, high elongation and triangularity. It mainly consists of superconducting magnets of poloidal and toroidal field （PF ＆ TF）, vacuum vessel （VV）, thermal radiation shield （TRS） and cryostat vessel （CV）. The significant difficulty for assembly of EAST is tight installation tolerances, which are in the order of several tenth of a millimeter. In particular, the alignment of plasma facing components to the magnetic axis of the device is less than ±0.5 mm. At present, a reasonable assembly process of EAST has been defined, and based on it, the alignment method for EAST, including the survey control network, the location of the main components in different directions, the magnetic axis determination and the accurate positioning of the plasma facing components inside of the vacuum vessel and so on, has been defined by using the sophisticated optical metrology system （SOMS）. This paper describes the assembly procedure of EAST and the installation tolerances associated with the main components. Meanwhile, how to establish the assembly survey control network, magnetic axis determination methods, are introduced in detail.