TheWendelstein 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 nonconformities, 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.

The goal of the Wendelstein 7-X (W7-X) stellarator project is to demonstrate that this type of machine is a viable option for a fusion power-plant. At present the W7-X experiment is in the assembly phase at the Max-Planck-Institut for plasma physics in Greifswald, Germany. The reliable prediction of the structural behavior of the W7-X machine is only possible by employing complex finite element (FE) analyses with a hierarchical set of FE models. A special strategy has been developed for the structural analysis which is under implementation now. This paper gives an overview of the analysis strategy, the applied structural criteria and critical issues, and focuses on the most remarkable results. The main attention is paid to the components that have been changed or optimized recently.

The Wendelstein 7-X (W7-X) stellarator project goal is to demonstrate that the stellarator is a viable option for a fusion power plant. W7-X is in an advanced state of construction and has entered the assembly phase at the Max-Planck-Institute fur Plasmaphysik (IPP) in Greifswald, Germany.The W7-X "pentagonal" basic magnet system is highly sensitive to parameter variations; the cryostat comprises two vessels, which are interconnected elastically by 299 ports. The strategy of the structural analysis for this complex mechanical system is being developed and implemented with the ultimate goal to create a tree of numerical models which reliably predict the stellarator structural behaviour.This paper gives an overview of the strategy, addresses the critical issues and focuses on the most interesting results of the analyses. (C) 2007 Elsevier B.V. All rights reserved.

The main parameters of W7-X are shown in table 1 [2,3]. The magnet system consists of 50 superconducting Non-Planar-Coils (NPC), 20 superconducting Planar Coils (PC) and the mechanical structure, which is based on the Central Ring and the intercoil support structure. The NPC and the PC are supported by the Central Ring through the Central Support (CS) elements, two for each coil (Fig.1). The Narrow Supports (NS) and the Lateral Supports (LS) connecting adjacent NPC casings in the inner and outer region of the machine respectively (Fig.1) and the Planar Supports connecting the PC to the NPC are the elements of the intercoil support structure. The coils are arranged toroidally in five equal modules, each one consisting of two flip symmetric semi-modules. One semi-module includes 5 differently shaped NPCs and 2 PCs. The Plasma Vessel (PV) has to closely follow the twisted shape of the plasma and it is

The Wendelstein 7-X (W7-X) stellarator project goal is to demonstrate that the stellarator is a viable option for a fusion power-plant. The W7-X is in advanced construction phase and has entered the assembly phase in Greifswald, Germany. The W7-X "pentagonal" basic magnet system is highly sensitive to the parameter variations; therefore the strategy of the structural analysis with the ultimate goal to create a tree of numerical models, which has to reliably predict stellarator structural behaviour, is in development and implementation phases. This paper gives an overview of the strategy and focuses on challenging boundary conditions and the most interesting results of the analyses

The large-scale mechanical response of the JET Vacuum Vessel and other structural components to disruption-induced loads has been analysed. The observed toroidal variations of the VDE-induced rocking motion can be explained by non-axisymmetric plasma forces. New evidence has been found in support of a plasma kink mode model which could account for the generation of the observed sideways forces.

Low cycle/high strain fatigue tests at 300°C have been carried out on specimens of Inconel 600 to assess the life of critical parts of the JET (Joint European Torus) vacuum vessel. Specimens have been loaded with alternate cycling strain up to ±1%. The shape of some specimens has been chosen to reproduce the real working conditions and the stress distributions. The results of the tests have been used to evaluate the actual damage produced by plasma disruptions in the past operational campaigns and to predict the additional damage caused by the expected future operational phases of JET. The critical regions of stresses and the maximum strain value for reference disruption scenarios have been evaluated. A statistical analysis of the measured vessel displacements for the past four years of operation has been carried out. It shows that the accumulated fatigue damage in the critical part is no more than a few percent. Anticipated total damage for another five years is less than 1/3 of the fatigue failure obtained from the tests

JET was designed for a plasma current of 5 MA and has operated successfully at that level. To enable JET to produce meaningful DT plasmas, it is necessary to upgrade the machine performance.1 The paper describes the effects on the poloidal and toroidal magnet systems of increasing the plasma current to 7 MA. It has not been necessary to increase the toroidal field but operation at higher plasma current increases the torque loading on the coils. In the case of the poloidal coils an increased flux swing is required so the magnetising current has been increased by 50%. Effects considered include magnetic forces and mechanical and thermal stresses in the coils. Modifications to the coil system and improvements to the power supplies that enable the new performance to be achieved are described. It is concluded that a 7 MA plasma current is feasible.

Disruptions at high current cause large forces and stresses at the JET vessel. Particularly undesirable are vertical disruptions due to possible failure of the vertical stabilisation system. Forces and stresses are being assessed on the basis of magnetic measurements, vessel deflections, strain gauge measurements at vertical supports and by a finite element mechanical analysis. The present operating restrictions may be relaxed after implementation of two support rings at the inner periphery of the vessel which supplement those at the outer periphery and will prevent unacceptable radial deflections.

Design details of the new in-vessel components are described with emphasis on those details which ensure: optimum operational as well as installation and remote handling capabilities, precise geometry and high reliability in severe environmental conditions in terms of heat load, forces and radiation.

Utilising to the fullest the enhanced capabilities of JET requires the installation of new in- vessel components. In this paper we discuss the requirements and describe the hardware solutions which are to be implemented and their effect on the future operation.

Two large vertical shaft flywheel generators each provides the JET (Joint European Torus) device with peak power up to 400 MW and energy up to 2600 MJ per pulse to induce and confine the multi-mega-ampère plasma current. The integrated rotor flywheel consists of a 650 tonne/10 m diameter rim carrying the poles of the machine. The energy is stored kinetically during a 9 min interval of acceleration from half-speed to full-speed and then released during a 20 s long deceleration. A design life of 100 000 cycles at full energy rating was specified. Following a review of the mechanical design and construction of the generators, the paper will concentrate on the assessment of the stresses and fatigue life of the rotor system, its dynamic behaviour (rim movement, critical speed and balancing) and on the performance in operation of the large thrust bearing.

The general design of the Joint European Torus (JET) is briefly described. The loads on its major structural components, at normal operation, and in cases of plasma instability and/or disruption, are discussed. The way these components have been assessed and optimised in relation to their loads is presented. A short account of mechanical design problems of auxiliary equipment is given. Finally, the state of operation of JET and its implications for the mechanical design is summarized. The mechanically most important components of the JET device are the support structure of the toroidal magnet, the vacuum vessel, the coils of the magnets and the pedestals supporting the weight of the torus. These components all participate in resisting and transmitting the primary forces during operation.

Heating the JET plasma well above temperatures reached in the ohmic phase is the aim of the two additional heating systems planned for JET: ion cyclotron resonance heating (ICRF) and neutral beam injection (NBI). Operations with the latter started in February 1986, initially with hydrogen injection, up to a power level of 7 MW. ICRF power has been delivered to the plasma by three antennae and has reached power levels of 6 MW for 2 s. In most experiments, the frequency of the waves is adjusted to position the minority ion resonance layer close to the centre of the discharge, resulting in a centrally peaked power deposition profile. Results have also shown that the increase of the volume averaged electron temperature was much less dependent on the radial position of the ion resonance layer than were the central temperatures of both ions and electrons which can both approach 5 keV. In cases of 'on axis' heating, large increases in the sawtooth activity is observed with sawtooth periods exceeding 0.3 s. Initial results with NBI have shown a decrease of the global energy confinement time with additional power similar to the one observed with ICRF heating.

As a direct continuation of Part I, where the theoretical background for the rheological model was discussed, possible operations on the model are discussed and performed on anterior cruciate ligament preparations from rabbits. Different methods of evaluating model constants are compared and other mathematical expressions than those of the model proposed are tried and discussed. Model parts are verified and numerical values are given for certain constants.