Manufacture and test of the non-planar coils for WENDELSTEIN 7-X

The stellarator WENDELSTEIN 7-X, a superconducting fusion experiment, is presently under construction at the Greifswald branch of the Max-Planck-Institut für Plasmaphysik (IPP). The standard magnetic configuration of WENDELSTEIN 7-X (W7-X) is formed by 50 non-planar superconducting coils. Twenty additional planar superconducting coils can modify the magnetic configuration.This paper describes the production of the non-planar coils including the production of the winding packs, the cases and the assembly procedure of the coils. Five coils are already finished and tested at room temperature at the manufacturer's site.All coils will also be tested under cryogenic conditions. The main tests are at nominal current, a quench test by increasing the temperature of the helium, a high voltage test, a pressure and a helium leak test, the measurement of the shrinkage during cool down and the deformations due to electromagnetic forces. The test procedure and the results of the first tests of the coils are also presented.     

Finite Element Analyses and Instrumentation Layout for Single Coil Testing of TF Coils in HT-7U

The HT-7U tokamak is a magnetically-confined full superconducting fusion device, consisting of superconducting toroidal field (TF) coils and superconducting poloidal field (PF) coils. These coils are wound with cable-in-conductor (CICC) which is based on UNK NbTi wires made in Russian '. A single D-shaped toroidal field magnet coil will be tested for large and expensive magnets systems before assembling them in the toroidal configuration. This paper describes the layout of the instrumentation for a superconducting test facility based on the results of a finite element modeling of the single coil of toroidal magnetic field (TF) coils in HT-7U tokamak device. At the same time, the design of coil support structure in the test facility is particularly discussed in some detail.     

Cyclic testing of shear keys for the ITER magnet system

Shear keys are to be used to support the out-of-plane loading of the toroidal field (TF) coils during a plasma pulse in ITER. At the inner intercoil structures (IIS) a set of poloidal shear keys is used to take the shear load at each connection between adjacent TF coils. Solid circular keys have been selected as reference. At the outer intercoil structures (OIS) adjustable conical shear keys and friction joint based shear panels are used to take the shear load. Low voltage electrical insulation is required at the flanges of the IIS and OIS, plus for all the bolts, poloidal keys and adjustable keys. This electrical insulation has to withstand large compression associated with some shear or slippage. A ceramic coating was selected for this purpose. The main scope of the experimental campaign was the mechanical testing of the shear keys and the electrical insulation in operational conditions relevant to ITER. Both keys were made of Inconel 718, provided with a ceramic alumina coating and inserted into flanges made of cast AISI 316 LN. The adjustable conical shear key was pre-loaded at room temperature and subject to cyclic shear loads of 2.5 MN for a large number of cycles (about 30,000) at cryogenic temperature (77 K). The conical key and the alumina coating remained undamaged after the test. Another test campaign was then performed with higher shear loads (up to 3 MN) to reach a sufficient safety margin even with the friction effect due to the pre-load. A set of 15,000 cycles were completed followed by some cycles at higher loads to reach the ultimate limit, which is the shear load to be experienced by the key in case of a poloidal field (PF) coil short.     

AC operation of large titanium sublimation pumps in a magnetic field: Results of the test stand for the W7-X neutral beam injectors

A neutral beam injection (NBI) system is being built for the Stellarator experiment Wendelstein 7-X (W7-X) currently under construction at IPP Greifswald. The NBI system consists of two injectors which are essentially a replica of the system present in the Tokamak experiment ASDEX-Upgrade at IPP Garching. A vacuum system with high pumping speed and large capacity is required to ensure proper vacuum conditions in the neutral beam line. For this purpose, large titanium sublimation pumps (TSP) are installed inside the NBI boxes, consisting of 4 m long hanging wires containing Ti and the surrounding condensation walls. The wires are DC ohmically heated up with 142 A to Ti sublimation temperature. A TSP system has been operated since many years in the AUG-NBI system, sublimating Ti in the pauses between the plasma discharges, when no magnetic field is present. However, at W7-X the superconducting coils generate a magnetic field permanently during experimental campaigns, whose stray B field with a maximum of 30 mT, affects the TSPs. Operated with DC, the wires would be deflected against the surrounding panels due to the Lorentz force. A simple possible solution is heating with AC, which reduces the wire deflection amplitude, inducing a risky wire oscillation. The feasibility of the AC operation in an equivalently strong B field such as the stray B field around W7-X has been demonstrated in a test stand for different AC waveforms and frequencies. Several test campaigns have shown no qualitative difference in the pumping properties between AC and DC operation of the TSP and no critical dynamic behaviour of the wires.     

EAST低温系统的故障树分析

EAST是世界上进行聚变研究的先进超导托卡马克实验装置.低温冷却系统是EAST的主要子系统之一,担负着纵向和极向场线圈、纵场线圈盒、冷屏和电流引线的冷却功能,需保证其能长期稳定运行.本文运用故障树分析法来评价低温系统的可靠性,使用自主开发的软件RiskA建立低温系统的故障树模型并进行定量计算,得到系统的失效概率和部件的重要度,并提出改进建议,为系统的优化设计和提高可靠性提供参考.     

Structural analysis and optimization for ITER upper ELM coil

ITER ELM coils are used to mitigate or suppress Edge Localized Modes (ELM), which are located between the vacuum vessel (VV) and shielding blanket modules and subject to high radiation levels, high temperature and high magnetic field. These coils shall have high heat transfer performance to avoid high thermal stress, sufficient strength and excellent fatigue to transport and bear the alternating electromagnetic force due to the combination of the high magnetic field and the AC current in the coil. Therefore these coils should be designed and analyzed to confirm the temperature distribution, strength and fatigue performance in the case of conservative assumption. To verify the design structural feasibility of the upper ELM coil under EM and thermal loads, thermal, static and fatigue structural analysis have been performed in detail using ANSYS. In addition, design optimization has been done to enhance the structural performance of the upper ELM coil.     

HFETR试验回路中考验装置热功率测量

利用数字仪表具有运算功能，将高通量工程试验堆试验回路中考虑装置热功率的有关函数关系输入到测量系统，实现了高温高压回路热功率的自动测量。为了满足不同试验对象的要求，设计了较大的功率测量范围，可以根据不同的试验工况定义仪表量程。     

Technical diagnosis system for EAST tokamak

Technical diagnosis system (TDS) is one of the important subsystems of EAST (experimental advanced superconducting tokamak) device, main function of which is to monitor status parameters in EAST device. Those status parameters include temperature of different positions of main components, resistance of each superconducting (SC) coils, joint resistance of SC coils and high-temperature superconducting (HTS) current leads, strain of cold-quality components endured force, and displacement and current of toroidal field (TF) coils in EAST device, which are analog input signals. In addition there are still some analog and digital output signals. The TDS monitors all of those signals in the period of EAST experiments. TDS data monitoring is described in detail for it plays important role during EAST campaign. And how to protect the SC magnet system during each plasma discharging is presented with data of temperature of coolant inlet and outlet of SC coils and feeders and cases of the TF coils and temperature in the upper and middle and bottom of the TF coil case.During construction of the TDS primary difficulties come from installation of Lakeshore Cernox temperature sensors, strain measurement of central solenoid coils support legs and installation of co-wound voltage sensors for quench detection. While during operation since the first commissioning big challenges are from temperature measurement changes in current leads and quench detection of PF coils. Those difficulties in both stages are introduced which are key to make the TDS reliable. Meanwhile analysis of experimental data like temperature as a back up to testify quench occurrence and stress on vacuum vessel thermal shield and vacuum vessel have also been discussed.     

Thermal analysis of the Mirnov coils of Wendelstein 7-X

Mirnov coils are used to measure fluctuations of the magnetic field which are in particular generated by magnetohydrodynamic (MHD) modes. The underlying plasma currents have a multipolar structure in a poloidal cross-section. Therefore the amplitude of the magnetic fluctuations decays quickly with increasing distance from the plasma edge. It is hence important to place the Mirnov coils as close to the plasma edge as possible where they are exposed to high thermal loads. Two types of Mirnov coils are proposed to be used in Wendelstein 7-X (W7-X). Type 1 (44 Mirnov coils) should be mounted on the plasma side of wall protection panels with a graphite cap to shield them from direct plasma exposure. Type 2 (137 Mirnov coils) will be located behind the tiles of the heat shields. An important issue concerning the design of these Mirnov coils is to verify their suitability for steady state operation from the thermal point of view. Both steady state and transient finite element thermal analyses were performed for the Mirnov coils under different conditions and with different designs. The paper presents detailed thermal analyses of the Mirnov coils.     

JT-60SA Toroidal Field Coils test cryostat development

Within the Broader Approach Agreement, Fusion for Energy will deliver to the Japanese Atomic Energy Association, amongst other components, the 18 Toroidal Field Coils (TFCs) for the superconducting Tokamak JT-60SA [1]. These coils will be individually tested at cryogenic temperatures and at the nominal current in a test cryostat. This cryostat is provided as an in-kind contribution by Belgium and is being developed jointly with CEA-Saclay/France.The vessel is large, oval shaped with an overall length of 11 m, a width of 7.2 m and a height of 6.5 m. To reduce the heat load to the coils the cryostat is covered by LN₂ cooled thermal shields. In addition to the cryostat, three test frames for the coils, the valve box vessel and the insulation vacuum system are also provided by Belgium. The Belgian contribution is design, manufacturing, assembly and test of the vacuum chamber, thermal shield and test frames by the Belgian company Ateliers de la Meuse (ALM), with the support of Centre Spatial de Liège (CSL). The TF coil test facility is assembled and the coil tests are performed by CEA/Saclay.The Belgian contribution, namely the design, manufacturing, assembly and test of the vacuum vessel, the thermal shields, and the test frames as well as of the vacuum pumping system are described in the presentation.     

A coil test facility for the cryogenic tests of the JT-60SA TF coils

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy (“Fusion for Energy”, F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature.The project is presently in the detailed design phase.     

Mechanical Analysis and Optimization of ITER Upper ELM Coil & Feeder

International thermonuclear experimental reactor （ITER） edge localized mode （ELM） coils are used to mitigate or suppress ELMs. The location of the coils in the vacuum vessel and behind the blankets exposes them to high radiation levels and high temperatures. The feeders provide the power and cooling water for ELM coils. They are located in the chinmey ports and experience lower radiation and temperature levels. These coils and feeders work in a high magnetic field environment and are subjected to alternating electromagnetic force due to the interaction between high magnetic field and alternating current （AC） current in the coils. They are also subjected to thermal stresses due to thermal expansion. Using the ITER upper ELM coil and feeder as an example, mechanical analyses are performed to verify and optimize the updated design to enhance their structural performance. The results show that the conductor, jacket and bracket can meet the static, fatigue and crack threshold criteria. The optimization indicates that adding chamfers to the bracket can reduce the high stress of the bracket, and removing two rails can reduce the peak reaction force on the two rails arising from thermal expansion.     

Seismic analysis of ITER fourth PF (Poloidal Field Coil) feeder

The ITER feeder systems connect the ITER magnet systems located inside the main cryostat to the cryo-plant, power-supply and control system interfaces outside the cryostat. The main purpose of the feeders is to convey the cryogenic supply and electrical power to the coils as well as house the instrumentation wiring. The PF busbar which carries 52 kA current will suffer from high Lorentz force due to the background magnetic field inspired by the coils and the self-field between every pair of busbars. Except their mechanical strength and thermal insulation performance must be achieved, the dynamic mechanism on PF structure should be assessed. This paper presents the simulation and seismic analysis on ITER 4th PF feeder including the Coil Terminal Box and S-bend Box (CTB and SBB), the Cryostat Feed-through (CFT), the In-Cryostat-Feeder (ICF), especially for the ground supports and main outer-tube firstly. This analysis aims to study seismic resistance on system design under local seismograms with floor response spectrum, the structural response vibration mode and response duration results of displacement, membrane stress, and bending stress on structure under different directions actuating signals were obtained by using the single-seismic spectrum analysis and Dead Weight analysis respectively. Based on the simulative and analytical results, the system seismic resistance and the integrity of the support structure in the 4th PF feeder have been studied and the detail design confirmed.     

Magnet design considerations for Tokamak fusion reactors

Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors.     

The magnetic field configuration measurement on EAST tokamak

The magnetic field configurations of poloidal field (PF) and toloidal field (TF) are the base of tokamak plasma operation. They are determined by the parameters such as positions and structures of PF and TF coils. Parameters of TF and PF coils of a new fully superconducting tokamak with non-circular cross-section EAST will change when the coils are cooled down from the ambient temperature to 4 K. Because of the cryogenic and refrigerator system, these parameters cannot be measured directly. Using magnetic probes signals, we measured and reconstructed magnetic field configuration of TF and PF coils. Parameters such as the positions of PF coils, the profile of the toloidal field in radial direction, the ripple and error field of toloidal field are obtained from the measurements.     

Design of Tokamak ELM Coil Support in High Nuclear Heat Environment

In Tokomak, the support of the ELM coil, which is close to the plasma and subject to high radiation level, high temperature and high magnetic field, is used to transport and bear the thermal load due to thermal expansion and the alternating electromagnetic force generated by high magnetic field and AC current in the coil. According to the feature of ITER ELM coil, the mechanical performance of rigid and flexible supports under different high nuclear heat levels is studied. Results show that flexible supports have more excellent performance in high nuclear heat condition than rigid supports. Concerning thermal and electromagnetic （EM） loads, optimized results further prove that flexible supports have better mechanical performance than rigid ones. Through these studies, reasonable support design can be provided for the ELM coils or similar coils in Tokamak based on the nuclear heat level.     

Feeder components and instrumentation for the JT-60SA magnet system

The modifying of the JT-60U magnet system to the superconducting coils is progressing as a satellite facility for ITER by both parties of Japanese government and European commission in the Broader Approach agreement. The magnet system requires current supplies of 25.7 kA for 18 TF coils and of 20 kA for 4 CS modules and 6 EF coils. The magnet system generates an average heat load of 3.2 kW at 4 K to the cryogenic system. The feeder components connected to the power supply provide current supply. The cooling pipes connected to the cryogenic system provide coolant supply. The instrumentation of the JT-60SA magnet system is used for its operation.     

New Development of VPI Process for Large Superconducting Coils

High vacuum is required for Vacuum Pressure Impregnation (VPI) process of large coils used in cryogenic. The defects such as dry spots and over rich resins should be minimized in large superconducting coils used. Both sealing problems associated with the mold and over rich resin problems are eliminated by using vacuum bag mold method with which we can simplify the design of vacuum mold.     

Present status and one upgrading method with a cold compressor of the EAST sub-cooling helium cryogenic system

Since 2006,the superconducting toroidal field (TF) coils of the Experimental Advanced Superconducting Tokomak (EAST) have been successfully cooled by supercritical helium at a temperature of 4.5 K and a pressure of 4 bara in eleven experiments.To obtain higher operating currents and magnetic fields it is necessary to lower the operating temperature of the TF coils.The EAST sub-cooling helium cryogenic system,with a warm oil ring pump (ORP),was tested twice in cool-down experiments,which made the TF coils operate at 3.8 K.However,the long term operational stability of the sub-cooling system cannot be guaranteed because of the ORP's poor mechanical and control performance.In this paper,the present status of the EAST subcooling helium cryogenic system is described,and then several cooling methods below 4.2 K and their merits are presented and analyzed.Finally,an upgrading method with a cold compressor for an EAST sub-cooling helium cryogenic system is proposed.The new process flow and thermodynamic calculation of the sub-cooling helium system,and the main parameters of the cold compressor,are also presented in detail.This work will provide a reference for the future upgrading of the sub-cooling helium system for higher operation parameters of the EAST device.     

Numerical simulation and experiment of error field measurement using luminous trace of electron beam in SST-1

In contrast to the earlier experiments conducted in other machines,here,in SST-1 the error field measurement experiment is performed with a filled gas pressure ～8×10~(-4) mbar which helped to create a luminescent toroidal beam of electron path originated due to impact excitation and guided by the toroidal magnetic field.Beam path deviations are observed and recorded from radial and top ports using visible range cameras.Such creation and detection of the electron beam path differs from the earlier works where the gun emitted electron beam deviation in ultrahigh vacuum was detected on a collector-grid/fluorescent screen.In the present experiment,large beam deviations were observed.Later investigation of the experimental set-up reveals existence of a possible source of radial electric field in between the source and the vacuum vessel which are separately grounded.Thus,to understand the observed phenomena,experiments are numerically modeled with deviated TF coil set,PF coil set and the electron source location.A particle tracing code is used to follow the electron path in the magnetic field generated by the coil set of interest.Simulation results suggest that the large deviation corresponds to the E×B drifts and not due to the large field errors.Toroidally averaged field errors of the SST-1 TF coils at toroidal field of B_0=15 kG are negligibly small～B_0×10~(-6) or less,which should not adversely affect the plasma performance.