Structural analysis of W7-X: Overview

The Wendelstein 7-X (W7-X) modular stellarator is in the assembly phase at the Max-Planck-Institut für Plasmaphysik (IPP) in Greifswald, Germany. The goal of the project is to demonstrate that this type of machine is a viable option for a fusion power-plant. The “pentagonal” magnet system of the machine encompasses 50 non-planar and 20 planar superconducting coils with sophisticated support structure. Structural reliability of components as well as resulting deformations and displacements during various modes of operation have to be considered not only for the magnet system but also throughout the whole cryostat whose main components are the plasma vessel, outer vessel, ports, and thermal insulation.A reliable prediction of the W7-X structural behaviour is only possible by employing complex finite element (FE) simulations with a hierarchical set of FE models. A special strategy has been developed and implemented for the task.The design is basically completed, main parameters are defined, and most of the W7-X components are manufactured. Therefore, the focus in the analysis is being shifted to the creation of parametric FE models which allow performing fast analyses of possible non-conformities, changes in the assembly procedure, and future exploration of operational limits.This paper gives an overview of the implemented analysis strategy, the applied safety margins, and focuses on the most remarkable results.     

Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.     

Status of Wendelstein 7-X construction

Wendelstein 7-X (W7-X) represents the continuation of fusion experiments of the stellarator type at the Max-Planck Institute for Plasma Physics (IPP). The aim of W7-X is to demonstrate the suitability for a fusion reactor of this alternative type of magnetically confined plasma experiment. W7-X is being built at Greifswald in the northeast of Germany. The size of device (725 tons, height of 5 m, diameter 16 m) and the superconductive magnet system distinguish W7-X from earlier stellarators at IPP. The paper provides a summary of the status of the main components, the mastering of the technical challenges during component acceptance testing and during machine assembly. Latest results of the assembly work are especially highlighted. The scope of the construction of W7-X was modified and additional acceleration measures were implemented to mitigate risks and delays. Some aspects of these changes are explained in this paper.     

Towards assembly completion and preparation of experimental campaigns of Wendelstein 7-X in the perspective of a path to a stellarator fusion power plant

The superconducting stellarator device Wendelstein 7-X, currently under construction, is the key device for the proof of stellarator optimization principles. To establish the optimized stellarator as a serious candidate for a fusion reactor, reactor-relevant dimensionless plasma parameters must be achieved in fully integrated steady-state scenarios. After more than 10 years of construction time, the completion of the device is now approaching rapidly (mid-2014). We discuss the most important lessons learned during the device assembly and first experiences with coming major work packages. Those are (a) assembly of about 2500 large, water-cooled, 3d-shaped in-vessel component elements; (b) assembly of in total 14 superconducting current leads, one pair for each coil type; and (c) assembly of the device periphery including diagnostics and heating systems. In the second part we report on the present status of planning for the first operation phase (5–10 s discharge duration at 8 MW heating power), the completion and hardening of the device for full power steady-state operation, and the second operation phase (up to 30 min discharge duration at 10 MW heating power). It is the ultimate goal of operation phase one to develop credible and robust discharge scenarios for the high-power steady-state operation phase two. Beyond the improved equilibrium, confinement, and stability properties owing to stellarator optimization, this requires density control, impurity control, edge iota control as well as high density microwave heating. Of paramount importance is the operation of the island divertor, which is realized in the first operation phase as an inertially cooled conventional graphite target divertor. It will be replaced later on by the steady-state capable island divertor with its water-cooled carbon fiber reinforced carbon target elements.     

Manufacturing of the Wendelstein 7-X divertor and wall protection

The in-vessel components of Wendelstein 7-X (W7-X) with a total surface of 265 m² comprise the divertor and the wall protection. The high heat flux (HHF) and lower heat flux (LHF) target, the baffle, the end plates closing the divertor chamber, a cryo vacuum pump (CVP) and a control coil form one divertor unit. Steel panels and the graphite heat shield protect the wall, including the ports. The HHF target elements, the steel panels and the control coils are manufactured by industry. The remaining components will be manufactured by the Max-Planck-Institute für Plasmaphysik (IPP) at its Garching workshops. For all components the final acceptance tests will be performed by IPP. This paper summarizes the main aspects for manufacturing, the preceding development and qualification tests as well as the final acceptance tests for the in-vessel components.     

Design principles for handmade electrical insulation of superconducting joints in W7-X

The superconducting magnet system of the Wendelstein 7-X (W7-X) experiment consists of 50 non-planar and 20 planar coils, 121 bus bars and 14 current leads. The connection between bus bars, coils and current leads will be provided by 198 joints. The joints have to be insulated manually during the assembly of the machine in constraint positions and a tight environment. In general the insulation is based on glass tapes impregnated with epoxy resin and special G10 insulating pieces embedded in the glass tape insulation. In critical areas Kapton^®-foils are embedded in the insulation. All types of insulation were qualified at mock-ups in a 1:1 model of the expected environment in W7-X. The qualification programme comprises thermal cycling between room temperature and 77 K and high voltage tests under air, under vacuum and under reduced pressure (Paschen test). The paper describes the main principles used for different types of handmade Paschen-tight insulations in W7-X and the visual and electrical tests during and after assembly.     

Asymmetry of W7-X magnet system introduced by torus assembly

The magnet system of the stellerator Wendelstein 7-X (W7-X) consists of 5 modules of 14 superconducting coils with complex 3D shape each. After manufacturing the coils and assembly of the modules on temporary stands, the position of each module on the machine base was successfully optimized to minimize the electromagnetic (EM) field asymmetry. This asymmetry originates from inevitable geometric deviations of the coils from the target shape due to manufacturing and assembly tolerances.However, new deviations were introduced after module optimization due to bolting the modules of the magnet system together to a torus, removing temporary supports and further loading of the machine base with weight of additional components.In this paper, the geometrical deviations along the centre line of the coil currents are assessed through detailed step-by-step non-linear finite element (FE) simulation of the assembly procedure of the complete torus. The model is evaluated against measured displacements and reaction forces monitored during consequent assembly steps. The results are being used to quantify the obtained field asymmetry and countermeasures to minimize it.     

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.     

Configuration Management for Wendelstein 7-X

A complex system like the large superconducting Wendelstein 7-X stellarator necessitates a dedicated organizational structure which assures permanent consistency between the requirements of its system specification and the performance attributes of all its components throughout its life time. This includes well-defined processes and centrally coordinated information structures. For this purposes the department Configuration Management (CM) has recently been established at W7-X. The detailed tasks of CM for W7-X are oriented along common CM standards and comprise configuration identification, change management, configuration status accounting and configuration verification. While the assembly of W7-X is proceeding some components are still under procurement or even under design. Thus design changes and non-conformances may have a direct impact on the assembly process. Highest priority has therefore been assigned to efficient control of change and non-conformance processes which might delay the assembly schedule.     

Design and technical status of the EU contribution to ITER

Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R&D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost.This paper provides an update of the design and technical status of EU contributions to ITER.     

Thermo-mechanical behavior of retro-reflector and resulting parallelism error of laser beams for Wendelstein 7-X interferometer

A 10 channels interferometer will be used in the Wendelstein 7-X (W7-X) for plasma density control and density profile tracking with laser beams passing through the plasma. Due to complex shape of non-planar modular coils and divertor structure, there are no large poloidally opposite ports on the plasma vessel (PV). Therefore 10 in-vessel Corner Cube Retro-reflectors (CCRs) will be used. The CCRs are integrated in the water cooled heat shield and exposed directly to thermal loads from plasma radiation. Thermo-mechanical issues are very important for the design of the CCR because deformation and flatness as well as mutual angles of the three reflecting surfaces would affect the parallelism of the laser beams and the functionality of the interferometer. Intensive work has been done to explore a suitable design for the CCR concerning thermo-mechanical behavior. Previous studies Ye et al. (2008, 2009) and Köppen et al. (2011) focused on structural optimization to decrease thermal stress in the reflecting plates under the thermal loads, and on computation and check of curvature radii of the deformed reflecting surfaces with the design criterion that the curvature radius must be bigger than 200 m. The paper presents detailed thermo-mechanical analysis of the current improved CCR under thermal loads and bolt preloads. The results of the thermo-mechanical analysis were used for the study of the resulting parallelism error of the laser beams with newly developed and more reasonable design criterion.     

Production management and quality assurance for the fabrication of the In-Vessel Components of the stellarator Wendelstein 7-X

The In-Vessel Components (IVC) of the stellarator Wendelstein 7-X consist of the divertor components and the first wall (FW) with their internal water cooling supply and a set of diagnostics. Due to the significant amount of different components, including many variants, a tool called Production Managing System (PMS) has been developed to organize the fabrication and the associated quality assurance. The PMS works by building a database containing the basic parts and assembly data, manufacturing and quality control plans, and available machine capacity. The creation of this database is based mainly on the parts lists, the manufacturing drawings, and details of the working flow organization. As a consequence of the learning process and technical adjustments during the design and manufacturing phase, the database needed to be permanently updated. Therefore an interface tool to optimize the data preparation has been developed. PMS has been demonstrated to be an efficient tool to support the IVC production activities providing reliable planning estimates, easily adaptable to problems encountered during the fabrication and provided a basis for the integration of quality assurance requirements.     

The application of leak before break concept to W7-X target module

Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental stellarator of the helias type fusion reactor currently being built in Greifswald, Germany. This experimental stellarator is a complex structure, such as nuclear power plants and high level of safety requirements should be used for structural integrity analysis. It is thus not possible to obtain simple solutions for general cases, therefore sophisticated methods are necessary for the analysis. Inside the Plasma Vessel (PV) of W7-X there is a number of different components such as pipes, divertors, baffles and targets. A guillotine failure of one component is very dangerous for structural integrity of surrounding components located in PV. For this reason it is very important to evaluate possibility to apply “leak before break” (LBB) concept for W7-X. The LBB concept is widely used in the nuclear industry to describe the idea that in the piping carrying the coolant of a power reactor a leak will occur before a catastrophic break will occurred. LBB allows to conduct the structural design without considering the loads due to postulated line breaks.The LBB analysis was made for the case when plasma vessel is operating in “baking” mode. “Baking” is the mode, when the cooling system is working as a warming system and it heats the plasma vessel structures up to 160 °C in order to release the absorbed gases from the surfaces and to pump them out of the plasma vessel before plasma operation.The LBB analysis was performed for most loaded component of target module. According to the results of the analysis it is possible to conclude that target module 1H fulfils the LBB requirements.     

First experiences with the new W7-X like control system at the WEGA stellarator

The new quality of the superconducting fusion device Wendelstein 7-X (W7-X) is its capability of steady state operation. Additionally the fusion device W7-X is a very complex technical system. The modular and strongly hierarchical control system has been designed to cope with these two requirements unique for fusion devices.To minimize the risks before commissioning the control and data acquisition system at W7-X it will be thoroughly tested in a prototype installation at the WEGA stellarator. WEGA is a classical stellarator which allows steady state plasma pulses at a magnetic field of 0.5 T. Despite its lesser complexity WEGA has the same main components, e.g. magnetic coil systems, ECRH, and diagnostics as W7-X and is therefore considered to be a suitable test-bed for the control system.The installation of the new W7-X like control and data acquisition system has been finished in March this year. Individual components of the control system have already been commissioned during the installation phase. In April final commissioning and testing of the complete system took place. First discharges fully controlled by the prototype control system have been realized.The contribution will focus on first discharges controlled by the new system. Furthermore it presents first experiences that will incorporate into the further development of the control system and the tools for planning, preparation, and realization of plasma discharges.     

The in-vessel components of the experiment WENDELSTEIN 7-X

The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components and the wall protection with its internal cooling supply. The main components of the open divertor are the vertical and horizontal target plates which form the pumping gap, the cryo-vacuum pumps and the control coils. The divertor volume is closed by graphite shielded baffle modules and with divertor closures. All these components are designed to be actively water-cooled. For the first commissioning phase planned in 2014, an inertial-cooled test divertor will be installed instead of the actively water-cooled high heat flux divertor. The wall protection consists of graphite-protected heat shields in the higher loaded areas and stainless steel panels in the lower loaded regions. The wall protection cooling circuits are connected through 80 supply-ports via so-called “plug-ins”. It is envisaged to protect the diagnostic ports by panel-type port-liners. Special graphite-shielded port liners are used on the diagnostic injector and the neutral beam injector ports. The in-vessel components are mainly manufactured and tested at the Max-Planck-Institute für Plasmaphysik in its Garching workshop. Panels, high heat flux target elements and control coils are delivered by industrial partners. Manufacturing of the KiP (“Komponenten im Plasmagefäß”) is in plan. Delivery of the components will be in time.     

Probe manipulators for Wendelstein 7-X and their interaction with the magnetic topology

Probe manipulators are a versatile addition to typical plasma edge diagnostics.Equipped with material samples they allow for detailed investigation of plasma–wall interaction processes,such as material erosion,deposition or impurity transport pathways.When combined with electrical probes,a study of scrape-off layer and plasma edge density,temperature and flow profiles as well as magnetic topologies is possible.A mid-plane manipulator is already in operation on Wendelstein 7-X.A system in the divertor region is currently under development.In the present paper we discuss the critical issue of heat and power loads,power redistribution and experimental access to the complex magnetic topology of Wendelstein 7-X.All the aforementioned aspects are of relevance for the design and operation of a probe manipulator in a device like Wendelstein?7-X.A focus is put on the topological region that is accessible for the different coil current configurations at Wendelstein 7-X and the power load on the manipulator with respect to the resulting different magnetic configurations.Qualitative analysis of power loads on plasma-facing components is performed using a numerical tracer particle diffusion tool provided via the Wendelstein 7-X Webservices.     

Electromagnetic load calculation of the ITER machine using a single finite element model including narrow slits of the in-vessel components

We evaluate electromagnetic (EM) loads on the main systems of the ITER machine using a single finite element model. The 20° sector of the full ITER machine includes the main in-vessel components as well as the vacuum vessel. Narrow slits of the in-vessel components are effectively modeled by using the element splitting method without significant increase of computation memory and time as well as without sacrificing the accuracy. Furthermore, the halo current is taken into account at the same time together with the plasma current. To apply both currents concurrently, dedicated conversion codes are utilized to transfer the plasma simulation results by DINA to the electromagnetic analysis by ANSYS-EMAG used here. The electromagnetic loads on the ITER machine are calculated for various disruption scenarios. Investigation on the analysis results is made to find the worst plasma disruption case and the design-driving load component for each system as well as to compare load contribution from eddy and halo currents. The effect of the narrow slits on load reduction is also examined.     

Thermo-mechanical analysis of Wendelstein 7-X plasma facing components

The stellarator experiment Wendelstein 7-X (W7-X) is designed for stationary plasma operation (30 min). Plasma facing components (PFCs) such as the divertor targets, baffles, heat shields and wall panels are being installed in the plasma vessel (PV) in order to protect it and other in-vessel components. The different PFCs will be exposed to different magnitude of heat loads in the range of 100 kW/m²–10 MW/m² during plasma operation. An important issue concerning the design of these PFCs is the thermo-mechanical analysis to verify their suitability for the specified operation phases. A series of finite element (FE) simulations has been performed to achieve this goal. Previous studies focused on the test divertor unit (TDU) and high heat flux (HHF) target elements. The paper presents detailed FE thermo-mechanical analyses of a prototype HHF target module, baffles, heat shields and wall panels, as well as benchmarking against tests.