Structural design of cryogenic component support for EAST （HT-7U）

EAST is a full superconducting tokamak with an elongated plasma cross-section. It consists of superconducting poloidal field (PF) magnet system, toroidal field (TF) magnet system, vacuum vessel with inner parts, thermal shields and cryostat vessel. The mission of the project is to widely investigate both physics and technologies of advanced tokamak operations, especially the mechanism of power and particle handling for steady-state operations. The cryogenic component is mainly composed of superconducting TF and superconducting PF coils that ensure the ability of sustaining magnetic field for plasma confinement, control and shaping in steady-state. This report describes the process of the structure design of cryogenic component support for EAST.     

Modeling of a confinement bypass accident with CONSEN,a fast-running code for safety analyses in fusion reactors

The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.     

Fusion Specific Features in ITER Accident Analysis

Fusion specific features like inherent plasma shutdown, low decay heat densities, cryogenic temperatures, and limited source terms were considered during the safety design process of ITER. Uncertainties in plasma disruptions motivates a robust design to cope with multiple failures of in-vessel cooling piping. A vacuum vessel pressure suppression system mitigates pressure transients and effectively captures mobilized radioactivity. In case of pump trips or ex-vessel coolant losses in the divertor the plasma needs to be actively terminated in a few seconds. Failure to do so might damage the divertor but radiological consequences will be minor due to the intact first confinement barrier. Tritium plant inventories are protected by several layers of confinement. Uncontrolled release of magnet energy will be prevented by design. Postulated damage from magnets to confinement barriers causes fluid ingress (air, water, helium) into the cryostat. The cold environment limits pressurization. Most tritium and dust is captured by condensation.     

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.     

Thermo hydraulic and quench propagation characteristics of SST-1 TF coil

SST-1 toroidal field (TF) magnet system is comprising of sixteen superconducting modified ‘D’ shaped TF coils. During single coil test campaigns spanning from June 10, 2010 till January 24, 2011; the electromagnetic, thermal hydraulic and mechanical performances of each TF magnet have been qualified at its respective nominal operating current of 10,000 A in either two-phase or supercritical helium cooling conditions. During the current charging experiments, few quenches have initiated either as a consequence of irrecoverable normal zones or being induced in some of the TF magnets. Quench evolution in the TF coils have been analyzed in detail in order to understand the thermal hydraulic and quench propagation characteristics of the SST-1 TF magnets. The same were also simulated using 1D code Gandalf. This paper elaborates the details of the analyses and the quench simulation results. A predictive quench propagation analysis of 16 assembled TF magnets system has also been reported in this paper.     

ITER重力支撑的制造设计、认证测试及关键技术研究

重力支撑（GS）作为国际热核聚变实验堆（ITER）磁体支撑系统的关键部件,不但要承受环向场超导磁体净重以及交变的电磁载荷,同时还需隔离来自杜瓦环的热量以维持环向场超导线圈的热稳定性。本文通过有限元分析和工程测试验证了GS结构设计的可靠性;通过换热分析和真空热交换效率测试验证了热锚连接结构的可靠性;通过全尺寸螺栓77 K疲劳测试验证了螺栓原型件的疲劳性能。在随后的制造过程中,使用液压拉伸器和研制的高精度螺栓伸长量测量装置对所有的螺栓进行了均匀、精确地紧固。真空正压氦检漏的测试结果证明了GS的泄漏率远低于ITER的要求。基于以上工程测试的结果,本文设计的GS的结构是可行的且能运用于ITER装置中。      

Preliminary design of KTX magnet system

KTX is a reversed field pinch magnetic confinement device of which the magnet system is designed in ASIPP and USTC. The main parameter of KTX is between RFX and MST. Its magnet system includes the toroidal field (TF) winding and poloidal field (PF) winding (ohmic heating winding and equilibrium field winding), which are less complex than tokamak device due to the fact that a tokamak requires a superconducting system to perform quasi-steady state operation and achieve Q > 10. However, the most important part of the magnet system design lies in how to keep the TF magnetic field ripple, as well as any kinds of stray field, to a minimum value. The main design activities of the KTX magnet system are presented as detailed as possible in this paper, and the main activities which have already been completed include magnet coils position and winding, insulation design, plasma modeling prediction, thermal analysis, magnetic field calculations were analyzed and so on. The magnet system design is one of the major activities for KTX device design, which is effective guarantee for the future R&D and manufacture. Besides, the detailed design activities should be continuously optimized and changed based on the results from future R&D and relevant tests.     

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.     

Operational experience with forced cooled superconducting magnets

Force-cooled concept has been chosen for ITER superconducting magnet to get reliable coil insulation using vacuum-pressure impregnation (VPI) technology. However 17 breakdowns occurred during operation of six magnets of this type or their single coil tests at operating voltage < 3 kV, while ITER needs 12 kV. All the breakdowns started on electric, cryogenic and diagnostic communications (ECDCs) by the high voltage induced at fast current variations in magnets concurrently with vacuum deterioration, but never on the coils, though sometimes the latter were damaged too. It suggests that simple wrap insulation currently employed on ECDCs and planned to be used in ITER is unacceptable. Upgrade of the ECDC insulation to the same level as on the coils is evidently needed. This could be done by covering each one from ECDCs with vacuum-tight grounded stainless steel casings filled up with solid insulator using VPI-technology. Such an insulation will be insensitive to in-cryostat conditions, excluding helium leaks and considerably simplifying the tests thus allowing saving time and cost. However it is not accepted in ITER design yet. So guarantee of breakdown prevention is not available.     

EBT reactor magnetics and particle confinement

Optimization of the vacuum magnetic field of an ELMO Bumpy Torus (EBT) reactor is investigated. Several methods of improving reactor volume utilization and single particle confinement are analyzed. These include the use of (a) a large number of sectors and/or a large mirror ratio, (b) high field Nb₃Sn or Nb₃Sn/NbTi hybrid mirror coils, (c) split-wedge mirror coils, (d) axis-encircling aspect ratio enhancement (ARE) coils, and (e) recently developed field symmetrizing (SYM) coils. Of these, particle drift orbit and three-dimensional tensor pressure equilibrium calculations show that the use of SYM coils in conjunction with high field mirror magnets offers the most promise of good plasma performance in reactors that are smaller (by up to 50%) than previous reference designs that did not employ supplementary coils. Aspect ratio enhancement coils also offer an attractive alternative for improved confinement, but they do not have many of the advantages of SYM coils, particularly for reactor applications. Split-wedge mirror coils improve volume utilization and trapped particle confinement, but they do not enhance the confinement of transitional and passing particles. High field magnets improve confinement by permitting a larger mirror ratio and a larger plasma radius by virtue of their smaller cross-sectional area and higher current density. The relative merits of each magnetics configuration are discussed, including the effects on single particle confinement, reactor volume utilization, materials requirements, engineering design considerations, and reactor assembly, maintenance, and accessibility.     

Design and implementation of quench detection instrumentation for TF magnet system of SST-1

Steady State Superconducting Tokamak-1 (SST-1) at Institute for Plasma Research (IPR), India is now in engineering validation phase. The assembled Toroidal Field (TF) magnet system of SST-1 will be operated at 10 kA of nominal current at helium cooled condition of 4.5 K. A reliable and fail proof quench detection (QD) system is essential for the safety and the investment protection requirements of the magnets. This QD system needs to continuously monitor all the superconducting coils, which include 16 TF magnets, return-loop, bus bars and current leads. In case of any event initiating the normal resistive zone and reaching thermal run-away, the QD system needs to trigger the magnet protection circuits. Precision instrumentation and control system with 204 signal channels had been developed for detection of quench anywhere in the entire TF magnet system. In the present configuration of quench detection scheme, the voltage drop across each double pancake (DP) of each TF coil are compared with its two adjacent DPs for the detection of normal zone and cancelation of inductive couples. Two identical redundant systems with one out of two configurations are successfully commissioned and tested at IPR. This paper describes the design and implementation of the QD system, Installation experience, validation test and initial results from the recent SST-1 magnet system charging.     

Non-superconducting magnet structures for near-term, large fusion experimental devices

Water cooled copper magnets provide a means of producing high magnetic fields for tokamaks using a well developed existing technology. The basic function of these magnets is to provide reliable, both time varying and steady state, magnetic fields. Copper electrical properties, insulation, and water cooling systems play major roles in design selection. Aside from being electro-magnetic devices, coils designed for tokamaks must be self-supporting structures, capable of resisting large I × B magnetic forces. These magnets require the integration of both electrical and structural design considerations.Magnet integrity is enhanced by the presence of structures which lend additional external support. These external structural systems are highly stressed and, often, deflection limited.This paper describes the magnet and structural design in the following American tokamak devices: the Princeton Large Torus (PLT), the Princeton Divertor Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The Joint European Torus (JET), also presented herein, has a magnet structure evolved from several European programs and, like TFTR, represents state of the art magnet and structure design.The PLT device was designed in 1971 as a high plasma current tokamak. At the time it incorporated the latest in copper magnet and structure technology. Design features on this machine have in some fashion subsequently been incorporated on every major device built within the tokamak fusion community.     

Manufacturing and assembly of the plasma- and outer vessel of the cryostat for Wendelstein 7-X

Wendelstein 7-X is an advanced helical stellarator, which is presently under construction at the Greifswald branch of IPP. A set of 70 superconducting coils arranged in five modules provides a twisted shaped magnetic cage for the plasma and allows steady state operation. Operation of the magnet system at cryogenic temperatures requires a cryostat which provides thermal protection and gives access to the plasma. The main components of the cryostat are the plasma vessel, the outer vessel, the ports, and the thermal insulation. The German company, MAN Diesel & Turbo SE Deggendorf (former MAN DWE GmbH Deggendorf), is responsible for the manufacture and assembly of the plasma vessel, the outer vessel and the thermal insulation. This paper describes the manufacturing and assembly technology of the plasma and outer vessel of the cryostat for Wendelstein 7-X.     

Status of the ITER magnets

The first 2 years of the ITER IO has seen substantial progress towards the construction of the magnets, in three main areas. Firstly, the design has been developed under the conflicting constraints to minimise construction costs and to maximise plasma physics performance. Building construction momentum while updating the design to take account of new physics assessments of the coil requirements has been challenging. Secondly, with a stabilising design, it has been possible for the Domestic Agencies to launch the first industrial procurement contracts. And thirdly, critical R&D to confirm the performance of the Nb3Sn cable in conduit design is proceeding successfully.The design consolidation has been accompanied by design reviews involving the international community. The reviews conducted by magnet experts have enabled a consensus to be built on choosing between some of the design options in the original ITER basic design in 2001. The major design decisions were to maintain the circular Nb₃Sn conductor embedded in radial plates for the toroidal field (TF) coils and to maintain NbTi-based conductors for the PF coils. Cold testing, at low current, is also being introduced for quality control purposes for all coils.     

Overview of main-mechanical-components and critical manufacturing aspects of the Wendelstein 7-X cryostat

Wendelstein 7-X (W7-X) will be the world's largest superconducting helical advanced stellarator. This stellarator concept is deemed to be a desirable alternative for a future power plant like DEMO. The main advance of the static plasma is caused by the three dimensional shape of some of the main mechanical component inside the cryostat. The geometry of the plasma vessel is formed around the three dimensional shape of the plasma. The coils and their support structure are enclosed within the outer vessel. The space between the outer, the plasma vessel and the ports is called cryostat because the vacuum inside provides thermal insulation of the magnet system which is cooled down to 4 K. Due to the different thermal movements of both vessels and the support structure have to be supported separately. 10 cryo legs will bear the coil support structure. The plasma vessel supporting system is divided into two separate systems, allowing horizontal and vertical adjustments. This paper aims to give an overview of the main mechanical components of the cryostat. The authors delineate some disparate and special problems during the manufacturing of the components at the companies in Europe. It describes the current manufacturing and assembly.     

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.     

The design and the manufacturing process of the superconducting toroidal field magnet system for EAST device

The toroidal field (TF) magnet system of EAST (HT-7U), which consists of 16 superconducting coils enclosed in steel cases, has been manufactured to generate the magnetic field of 3.5 T at the plasma center to maintain plasma in a tokamak configuration with a current up to 1 MA. The TF coils have an approximately D-shape geometry of 2.6 m wide and 4.0 m high, and operate at a maximum field of 5.8 T. The conductor used in the TF coil is NbTi/Cu cable-in conduit (CIC) conductor, and its operating current is 14.3 kA.In March 2006, the first cooling down of the EAST device has been carried out successfully. The total of TF magnet system has been cooled down from room temperature to 4.5 K, and the TF system has been energized up to 8.2 kA with 5 A/s ramp rate. In September 2006, full performances of the TF magnet system have been reached, and the device of EAST has delivered its first plasma. In addition, the TF magnet system has been routinely operated with a current maintained constant on a whole day basis, for a preliminary program of more than 500 shots.In this paper, the main parts of the design, developmental tests, and the fabrication and assembly of TF coils are described in detail.     

Design and assembly technology for the thermal insulation of the W7-X cryostat

The Max-Planck-Institut für Plasmaphysik in Greifswald is building up the stellarator fusion experiment Wendelstein 7-X (W7-X). To operate the superconducting magnet system the vacuum and the cold structures are protected by a thermal insulated cryostat. The plasma vessel forms the inner cryostat wall, the outer wall is realised by a thermal insulated outer vessel. In addition 254 thermal insulated ports are fed through the cryogenic vacuum to allow the access to the plasma vessel for heating systems, supply lines or plasma diagnostics.The thermal insulation is being manufactured and assembled by MAN Diesel & Turbo SE (Germany). It consists of a multi-layer insulation (MLI) made of aluminized Kapton with a silk like fibreglass spacer and a thermal shield covering the inner cryostat surfaces. The shield on the plasma vessel is made of fibreglass reinforced epoxy resin with integrated copper meshes. The outer vessel insulation is made of brass panels with an average size of 3.3 × 2.0 m². Cooling loops made of stainless steel are connected via copper strips to the brass panels. Especially the complex 3 D shape of the plasma vessel, the restricted space inside the cryostat and the consideration of the operational component movements influenced the design work heavily. The manufacturing and the assembly has to fulfil stringent geometrical tolerances e.g. for the outer vessel panels +3/?2 mm.     

国际热核聚变实验堆过渡馈线设计与分析

国际热核聚变实验堆（InternationalThermonuclearExperimentalReactor，简称ITER）采用超导馈线系统传输磁体线圈所需电流、冷却液及数据信号等。本文介绍了馈线系统中过渡馈线的设计，并通过有限元法对设计结构进行全面校核和应力分析，为结构的进一步研制提供了理论依据。