Electrical Parameters of the Vacuum Vessel in HT-7U Tokamak

A method is presented to express the electrical parameters of the vacuum vessel in this paper. According to the results of numerical computation and the distribution of the eddy currents, the mutual inductance can be given by calculating the flux produced by the toroidal eddy currents. The time constants of the vacuum vessel of HT-7U tokamak are derived from the decay characteristics of the eddy currents.

Design of the Magnetic Field for HT-7U Tokamak

The HT-7U superconducting tokamak is an advanced steady-state plasma physics experimental device to be built in the Institute of Plasma Physics, the Chinese Academy of Sciences (IPP-CAS). The plasma current is 1 MA and the major and minor radius are 1.78 m and 0.4m respectively, with an elongation of 1.85. The preliminary engineering design of the poloidal field (PF) and toroidal field (TF) magnet systems have been done. The PF system is composed of twelve superconducting coils located symmetrically about the equator plane. The central solenoid (CS) assembly is formed by six coils. The TF system consists of 16 superconducting coils. The NbTi cable-in-conduit conductor or (CICC) cooled by a supercritical helium at 4.5 K is chosen as a superconductor for all of the PF and TF coils. At this temperature, the peak magnetic field on the PF magnets is about 4.51 T .The maximum volt-second capacity and the duration of plasma inductive discharge are about 10 Vs and 10 seconds respectively. The stray field in plasma initial region is quite low ( 1.5× 10 -3 T). The magnetic field on the TF magnet is 5.8 T while the toroidal field at the center of the device (R = 1.7 m) is 3.5 T and the ripple of the TF is less than 0.62% at the outer plasma surface (R = 2.1 m). All of the PF and TF magnets are stable during all modes of operation including the plasma disruption. The final design of the PF system is the result of an iterative process involving the use of equilibrium code EQT, magnetic code EFFI, and other codes, which have been developed by our designing group.

New CICC Design Versions with High Copper Fraction for HT-7U Tokamak

The peak fields of HT-7U TF/PF coils are designed at 5.8 T and 4.5 T, respectively. The superconducting (SC) NbTi/Cu strands for the coils have a quite low copper fraction (0.58). To increase the ratio of the limiting current Vs. critical current for the cable in conduit conductor (CICC), the first-stage subcable consists of several copper strands twisted around a SC composite one. The copper strand diameter is different from the SC one. Based on Bottura's CICC design approach we take the product of total copper cross-section and wetted perimeter as one of the key design parameters to meet with stability margin requirement. The product reaches 114 cm3 for 15.5 kA TF CICC at operating temperature of 4.2 K and 139 cm3 for 15 kA PF CICC at 3.8 K to obtain stability margins of 280 and 700 mJ/cm3, respectively.

Preliminary Design of Control Network for HT-7U Tokamak Cryogenic System

In the course of the cryoplant modernization, a control network will be set up in order to facilitate the control, the supervision, the centralized data acquisition and the alarm handling of the cryogenic system for HT-7U tokamak. The paper introduces the preliminary design of control network based on the Controller Link Network for HT-7U tokamak cryogenic system. The multi-layer structure mentioned in the paper is the mainstream of automatic control. The control philosophy, the structure of the network and the components for control are also presented.

Recent Progress of HT-7U Superconducting Tokamak

HT-7U is a superconducting tokamak, which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak.The engineering design of the device has been optimized. The RfcD program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.

Preliminary Design of Poloildal Field Power Supply for HT-7U Super-conducting Tokamak

This paper presents the preliminary design of poloidal field power supply system of HT-7U super-conducting tokamak. With an emphasis on AC/DC power converter, DC circuit breaker, quench protection, harmonic suppression and static var. compensation, and AC power system, the design principle and features are introduced, the design scheme and R & D progress are described, the simulation studies and laboratory test are presented too.

The Present Status of HT-7U Superconducting Tokamak Project

The first Chinese superconducting tokamak HT-7 was reconstructed on the basis of a Russian tokamak T-7.The main purpose of reconstruction is to improve the accessibility of the device and to provide a possibility of long pulse operation with high performance. The reconstruction has been done successfully. The HT-7U project has been approved and funded as a National Project, the engineering design and R & D are under way.

Reactive Power Compensation and Harmonic Suppression for Power Supply System of HT-7U Superconductive Tokamak

In this paper, a strategy for the reactive power compensation and harmonic suppression of the power supply system in HT-7U superconductive Tokamak is proposed. The optimized approach is given in the parameters design for passive filter. Also a controlling method with fast response time and good accuracy is put forward for the compensator, which is more suitable for the dynamic load.

Preliminary Engineering Design of Toroidal Field Magnet System for Superconducting Tokamak HT-7U

HT7U is a large fusion experimental device. It will be built in the Institute of Plasma Physics of Chinese Academy of Sciences. The mission of HT-7U is to develop the scientific basis for a continuously operating tokama-k fusion reactor. This paper describes only a toroidal field (TF) superconducting magnet system of HT7U. In this paper, design criteria of conductor and stability analysis, coil winding and support structure design of magnet system, mechanical calculation and stress analysis, heat load evaluation are given.

Development of Special Winding Machine for HT-7U Superconducting Tokamak

A special winding machine with high accuracy has just been developed and applied to the construction of HT-7U Tokamak. It is one of the critical facilities for R & D of HT-7U construction. The machine mainly consists of five parts, including a CICC pay-off spool, a fourroller correcting assembly, a four-roller forming/bending assembly, a continuous winding structure and a CNC control system with three-axis AC servo motors. The facility is used for Cable in Conduit Conductor (CICC) magnet fabrication of HT-7U. The main requirements of the winding machine are: continuous winding to reduce joints inside the coils; pre-forming CICC conductor to avoid winding with tension; suitable for all TF & PF coils of various coil shapes and within the dimension limit; improving the configuration tolerance and the special flatness of the CICC conductor. This paper emphasizes on the design and fabrication of the special winding machine for HT-7U. Some analyses and techniques in winding process for trial D-shape coil are also presented.

Experimental Results of MHD Suppression with Modulated LHCD on the HT-7 Tokamak

Modulation of lower hybrid current drive was used successfully to suppress MHD activity. This was achieved in discharges with MHD m = 2 tearing modes during the discharge conditions Ip = 110 kA, Bt = 1.75 T, ne0 - 1.1 × 1013 cm-3. The delivering time of LHCD pulse is less then 30 μs. The amplitude, interval and the period of LHCD modulation pulse can be adjusted very conveniently. The modulation LHCD can be delivered very fast at any time during the discharge. The modulation LHCD period was always much shorter than the plasma resistive time (Tη ≈100 ms). So the profile of plasma current is changed much faster than the plasma resistive time. The different forms of LHCD modulating can be proved.

Soft x-ray Diode-array Diagnostic System for HT-7 Superconducting Tokamak

A diagnostic system of soft x-ray diode-array was set up for HT-7 superconducting tokamak. The system consists of two slot-aperture cameras and is capable of measuring the soft x-ray emission from the plasma on HT-7 device with a high resolution in space and a high response in time. Both cameras, located separately in a horizontal port and a vertical port each with thirty-seven detectors of An-Si surface-barrier diode (SBD) can view the same toroidal cross-section of the plasma from different poloidal chords. In this paper, the structure, principle and performance of the diagnostic system are described and some experimental results observed are presented.

H_α/ D_α Measurements Based on Photo Diode Array in the HT-7 Tokamak

Photo Diode Array (PDA) has been successfully applied in HT-7 tokamak experiments. The PDA system is almost free of electromagnetic interference from the machine. The system is compact and inexpensive, and it is convenient to be arranged in experiment. With the PDA system, the particle confinement time (τp) has been systematically investigated. The relations such as τp on the center-line-averaged electron density (ne), τp on plasma current (Ip), and τp, on the toroidal magnetic field (Bt) have been obtained. The particle confinement under the Ion Berstain Wave (IBW) Heating has also been measured and analyzed.

Experimental Research on Lower Hybrid Current Drive in Superconducting Tokamak HT-7

?Fundamental experiments on lower hybrid current drive (LHCD) have been undertaken on HT-7 superconducting tokamak. The experiments on LHCD efficiency reveal its depen- deuce on plasma density and the toroidal magnetic field. Furthermore, the experiments on HT-7 successfully demonstrate the ability for LHCD to sustain long pulse tokamak discharges, such as discharges with full non-inductive current drive for several seconds. The experimental study to improve plasma confinements by LHCD suggests that the improvement should be due to the change o f current profile. It has also been demonstrated by the experiments that the lower hybrid wave may lead to an enhanced ionization of particles in the region where the wave is deposited.

A Study on Main Breaker for Quench Protection of HT-7U Toroidal Superconducting Tokamak

This paper proposes a quench protection project of HT-7U toroidal superconducting tokamak through a forced commutation analysis of DC circuit breaker (DCCB) paralleling fuse. Based on the requirement of quench protection, main parameters are selected. Experimental results demonstrate the validity of this proposed project.

Superconducting Busline Design for HT-7U Tokamak

Superconducting (SC) tokamak HT-7U has seven pairs of buslines connecting toroidal/poloidal coils and the current leads. These SC buslines (SCBLs) share a common cryostat and are made of the cable in conduit conductors (CICCs) arranged as a decoupling configuration. In order to reduce the heat loads conducted from the seven current leads with a capacity of 15 kA during the magnets cooldown, the buslines with a much lower thermal conduction were employed in comparison with the current leads, and a special cooling loop was designed.

Fast Thermography for HT-7U Superconducting Tokamak

A 30 channel fast IR pyrometry array has been proposed for HT-7U superconduct- ing tokamak, which has a 0.5 μs time response, a 10 mm diameter spatial resolution and a 5℃ temperature resolution. The temperature range is from 250℃ to 1200℃. The two dimensional temperature profiles of the first wall during both major and minor disruptions could be measured with an accuracy of about 1% of measuring temperature, which is adequate for tokamak experi- ments. This offers a very useful tool for disruption study, especially for the research of divertor physics and edge heat flux on HT-7U superconducting tokamak.

A Study of High-Power Thyristor Switch of Poloidal-Field Power Supply System in HT-7U Super-Conductive Tokamak

Through theoretical analysis of thyristor switch, criterion of turn-off was derived for the design of thyristor switch. The expression of parameter design and its math model during the turn-off were deduced. The simulation and experiment have been accomplished to validate the analysis.

Exposure of W-TiC/Cu Functionally Graded Materials in the Edge Plasma of HT-7 Tokamak

Six-layered W-TiC/Cu functionally graded materials were fabricated by resistance sintering under ultra-high pressure and exposed in the edge plasma of HT-7 tokamak. Microstruc- ture morphologies show that the TiC particles distribute homogeneously in the W matrix, strength- ening the grain boundary, while gradient layers provide a good compositional transition from W- TiC to Cu. After about 360 shots in the HT-7 tokamak, clear surface modification can be observed after plasma exposure, and the addition of nano TiC particles is beneficial to the improvement of plasma loads resistance of W.

Measurement of Magnetic Island Width by Multi-Channel ECE Radiometer on HT-7 Tokamak

A 16-channel electron cyclotron emission （ECE） radiometer has been employed to observe the （m, n） = （2, 1） magnetic island structure on HT-7 tokamak, where m and n represent the poloidal and toroidal mode number respectively. The results indicate that the island width is about 7 cm when the magnetic island is saturated during the m/n = 2/1 mode. The location of resonance surface can be determined by plotting the contour of ECE relative fluctuation. This method could be applied to the HT-7 and EAST campaigns in the future for the research of neoclassical tearing modes （NTMs）.