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.     

The Optical Sensors in FM-1

In the PPL FM-1 Machine, the plasma containment volume is shaped, in part, by a levitated superconducting ring which is stabilized and positioned by servoed currents in stationary tilt and slide coils. The instantaneous position of the ring is sensed, in the required five degrees of freedom, by external non-contacting reflex optical sensors. Design and construction of these sensors is described and results of performance tests, both off and on the FM-1, are presented.     

Feedback Control of Plasma Current and Horizontal Position in HT-7

1. IntroductionA superconducting tokamak HT-7 has been estab-lished at ASIPP, Hefei, China. The machine.was de-signed to mainly investigate the reactor-relevant ls-sues, such as edvanced operation modes and plasmawall interastions in the near-steady-state condition.Its poloidal fie1d coils include ohmic heatlng coi1s'bias field coils' vertical field coils and horizontalfie1d coi1s (See Fig.1), being connected to indlvldualpower supplies which are all the thyristor--controlledrectifier unlt…     

Poloidal Field Control for the HT—7U Superconducting Tokamak

Controlling the poloidal field(PF) in the HT-7U superconducting tokamak is critical to the realization of the mission of advanced tokamak research.Plasma start-up,plasma position,shape,current control and plasma shape reconstruction have been performed as a part of its design process.The PF coils have been designed to produce a wide range of plasmas,Plasma start-up can be achieved for multiple conditions.Fast controlling coils for plasma position inside the vacuum vessel are sued for controloling the plasma vertical position on a short timescale.The PF coils control the plasma current and shape on a slower timescale,VXI(VME bus extensions for Instrumentation)Bus system and DSP(Digital Signal Processor is a basic unit of the feedback control system),the response time of which is about(2-4)ms.The basic unit of this system ,the shape-controlling algorithms of a few critical points on plasma boundary and real-time equilibrium fitting(RTEFIT)will be described in this paper.     

Present Status of the KSTAR Superconducting Magnet System Development

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC,the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation , the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.     

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.     

软件补偿自编码棒位指示系统

目前,压水堆核电站普遍采用线圈编码式棒位指示系统测量其控制棒位置,而在此基础上发展发展的自编码数字式棒位指示系统的探测器与前述系统的探测器均需补偿线圈来实现差动输出,从而提高输出信号的信噪比。本研究提出并从理论上证明后者可取消专设的补偿线圈而采用软件补偿,并据此设计了软件补偿自编码棒位探测器。该方案进一步简化了棒位指示系统的探测器结构,使其线圈数量少,轴向长度缩短,结构简单的优点得以在２００ＭＷ供     

Insulating process for HT-7U central solenoid model coils

The HT-7U superconducting Tokamak is a whole superconducting magnetically confined fusion device. The insulating system of its central solenoid coils is critical to its properties. In this paper the forming of the insulating system and the vacuum-pressure-impregnating (VPI) are introduced, and the whole insulating process is verified under the superconducting experiment condition.     

小型活性靶时间投影室性能研究

本文设计了使用64路一维读出条的小型活性靶时间投影室,并对其性能进行了测试,其气体室的灵敏体积为10 cm×10 cm×14 cm,通过连接2层厚气体电子倍增器进行电子放大。为改善场笼边缘的电场不均匀性,引入场笼环的设计结构。当场笼环加高压为-950 V时,测得α粒子沿漂移电场方向的径迹位置分辨小于02 mm,径迹角度分辨小于0.6°,时间分辨小于20 ns。活性靶时间投影室的工作气体为96%He+4%CO₂。实验中也观察到了清晰的α粒子在He气上的弹性散射事件。     

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.     

Quench detection of fast plasma events for the JT-60SA central solenoid

The JT-60 is planned to be modified to a full-superconducting tokamak referred to as the JT-60 Super Advanced (JT-60SA). The maximum temperature of the magnet during its quench might reach the temperature of higher than several hundreds Kelvin that will damage the superconducting magnet itself. The high precision quench detection system, therefore, is one of the key technologies in the superconducting magnet protection system.The pick-up coil method, which is using voltage taps to detect the normal voltage, is used for the quench detection of the JT-60SA superconducting magnet system. The disk-shaped pick-up coils are inserted in the central solenoid (CS) module to compensate the inductive voltage. In the previous study, the quench detection system requires a large number of pick-up coils. The reliability of quench detection system would be higher by simplifying the detection system such as reducing the number of pick-up coils. Simplifying the quench detection system is also important to reduce the total cost of the protection system. Hence the design method is improved by increasing optimizing parameters. The improved design method can reduce the number of pick-up coils without reducing the sensitivity of detection; consequently the protection system can be designed with higher reliability and lower cost. The applicability of the disk-shaped pick-up coil for quench detection system is evaluated by the two dimensional analysis. In the previous study, however, the analysis model only took into account the CS, EF (equilibrium field) coils and plasma. Therefore, applicability of the disk-shaped pick-up coil for the quench detection system remains open question because the fast plasma events, such as disruption, mini collapse and ELM (edge localized mode), directly influences on the voltage of pick-up coil making the quench signal undetectable. Consequently, a new analysis model proposed in the present paper was designed to avoid this difficulty by introducing the passive coil series such as vacuum vessel and stabilizer. The influence of fast plasma events is absorbed by passive coil series like real system, and the evaluation of applicability can be examined in detail. The analysis results show that the disk-shaped pick-up coil is applicable whenever the standard operation, disruption, mini collapse and ELM.     

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.     

The design of the Tokamak Physics Experiment (TPX)

The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.     

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.     

Assessment of cracks in lateral supports of the magnet system of Wendelstein 7-X

The superconducting coils of the magnet system of Wendelstein 7-X (W7-X) are bolted onto a central support ring and interconnected with five so-called lateral support elements (LSEs) per half module. After welding of the LSE hollow boxes to the coil cases cracks were found in the vicinity of the welds that could potentially limit the allowed number N of electromagnetic (EM) load cycles of the machine.In response to the appearance of first cracks during assembly, the stress intensity factor (SIF) of theoretical cracks of various sizes in potentially critical position and orientation were predicted in a fast approach. For each crack size, N was based on the SIF, derived from beam theory, and on Paris’ law parameters determined in fatigue crack growth rate (FCGR) tests, thus leading to tolerable maximal crack sizes and distances between cracks. It was proved that the actual crack dimensions remained below these values or turned out to be only superficial. Afterwards, (extended) finite element method (XFEM and FEM) and boundary element method (BEM) models were developed to project the SIF of most critical tolerated cracks, considering new FCGR tests and the local stress state in more detail. N appeared highly sensitive to the assumptions which were therefore critically reviewed.Finally, the limit for load combinations of different amplitudes was determined using Miner's rule. As a result it was shown that the predefined number of W7-X operation cycles is not jeopardized by any of the detected cracks.     

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.     

Floating Ring Stabilization System

The necessity of magnetically levitating a superconducting, current carrying ring without mechanical supports requires an electrical stabilizing system. The ring, when supported, has only one equilibrium position and this position is one of unstable equilibrium. The ring is free to move with up to six degrees of freedom. Two degrees are dictated by the ability of the ring to move in any direction in the plane of the ring. Additionally, the ring can be tilted requiring three more reference points to determine a new plane. The sixth degree of freedom is the ability of the ring to rotate upon its own axis, but because external forces tending to cause this rotation can, in principal, be eliminated, a restoring supply is not provided. Each of the five identical stabilizing supplies in this system has a feedback loop composed of the following segments. The position detector element is an optical head which sends a light beam across an optical edge on the ring, the reflection of which is sensed on a photo diode. This diode voltage signal, proportional to the ring position, is the input to an amplifier whose output is a signal proportional to position, velocity and ring acceleration. A 50 volt, 100 ampere, bidirectional dc transistor power supply is activated from this signal which in turn powers the appropriate stabilizing coil located near the floating current carrying ring.     

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.     

Overcurrent analyses in JT-60SA poloidal circuits due to plasma disruption and quench protection intervention

The identification of the maximum amplitude of the currents circulating in the circuits is a useful indication for the design both of magnet and power supply components in fusion experiments. This paper evaluates the maximum level of coil overcurrents in the poloidal superconducting magnets of JT-60SA, the satellite tokamak that will be built in Naka, Japan, in the framework of EU-JA “Broader Approach” Agreement and that is expected to perform first plasma on 2016.To derive these information, a complete model capable to take into account all the mutually coupled elements was worked out, including the poloidal superconducting coils, the plasma position control in-vessel coils, the vacuum vessel, the stabilizing plates and the plasma.The model was utilized to analyze plasma disruption and quench protection circuit intervention in a large variety of different conditions to identify the possible overcurrent levels. The paper describes the model and the analyses performed, and presents and discusses the results.     

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.