Research and Development of Superconducting Magnet for Fusion Power

This paper describes mainly technological achievment of superconducting magnet for fusion power for the latest 10 years in Japan. The magnet development had been devoted to tokamak Fusion Experimental Reactor (FER). The major results obtained up to now are as follows.

In toroidal coil program, 12 T field generation, which is requested in a reactor toroidal coil, was realized with 6 kA multifilamentary Nb3Sn conductor in 1 m bore. For scaling-up of toroidal coil, half size coils of FER, LCT coils were tested up to 9 T.

In poloidal coil program, Demo Poloidal Coil project is now under way and coil testing will be started in spring of 1989. The stored energy of this coil is around 40 MJ.

In cryogenic technology program, fabrication and operation of large helium refrigerator technology were well established. A supercritical helium pump of 500 g/s was tested for forced flow coil.     

Plasma MHD generator for modelling energy conversion in pulsed thermonuclear reactors

Conclsuions The construction of an experimental model for studying MHD energy conversion from a pulsed thermonuclear reactor is a realistic technical task at the present time. Doing this would permit development of a large scale MHD generator module for the typical parameters of the heated working medium in a pulsed thermonuclear reactor.In principle it is possible to obtain an efficiency of at least about 40% with a linear plasma MHD generator. The efficiency of the whole plant might be increased further by utilization of the thermal energy at the outlet of the MHD channel in traditional methods.When such an MHD generator is built difficulties with the behavior of supersonic plasma streams undergoing strong velocity reduction in a channel and the associated gasdynamic problems can clearly be solved successfully by active modification of the boundary layer and appropriate profiling of the MHD channel. Some complications may arise if a regime with time varying magnetic braking is used. Also important is the problem of the behavior of the plasma stream at large magnetic Reynolds numbers (Re_m1).The basic technological problems are these: materials for the MHD channel, cooling arrangements for the channel (especially the critical cross section of the flow path), and pumping off the boundary layer at the electrodes and preventing lithium condensation on the channel walls. Because of the small magnetic field required, construction of the magnet system will clearly not present substantial technical difficulties associated with its size.The most important physical questions as well as a number of technological questions characteristic of this problem may be investigated on a fairly simple model MHD generator with an output power level of 300–500 MW, a pulse duration of 10–20 msec, and a lithium plasma source.Translated from Atomnaya Énergiya, Vol. 39, No. 6, pp. 387–391, December, 1975.     

How should we test the ITER TF coils?

The international thermonuclear experimental reactor (ITER) toroidal field (TF) magnet system consists of 18 superconducting coils using a 68 kA Nb₃Sn conductor. In order to guarantee the performances of these coils prior to their installation, the test of at least one prototype coil at liquid helium temperature and full current is required. The test of all coils in the two-coil test configuration, with successive charging of each coil to nominal current is recommended. This requires a large test facility.     

Magnet Safety Assessment for ITER

The ITER magnet system consists of structurally linked sets of toroidal (TF) and poloidal (PF) field coils, central solenoid (CS), and various support structures. The coils are superconducting, force flow Helium cooled with a Kapton-Glass-Epoxy multilayer insulation system. The stored magnetic energy is about 100GJ in the TF system and 20GJ in the PF-CS. Coils and structure are maintained at 4 K by enclosing them in a vacuum cryostat. The cryostat, comprising an outer envelope to the magnets, forms most of the second radioactivity confinement barrier. The inner primary barrier is formed by the vacuum vessel, its ports and their extensions. To keep the machine size within acceptable bounds, it is essential that the magnets are in close proximity to both of the nuclear confinement barriers. The objective of the magnet design is that, although local damage to one of the barriers may occur in very exceptional circumstances, large scale magnet structural or thermal failure leading to simultaneous breaching of both barriers is not credible. Magnet accidents fall into three categories: thermal (which includes arcing arising from insulation failure and local overheating due to discharge failure in the event of a superconductor quench), structural (which includes component mechanical failure arising from material inadequacies, design errors and exceptional force patterns arising from coil shorts or control failures), and fluid (Helium release due to cooling line failure). After a preliminary survey to select initial faults conceivable within the present design, these faults are systematically analyzed to provide an assessment of the damage potential. The results of this damage assessment together with an assessment of the reliability of the monitoring and protective systems, shows that the magnets can operate with the required safety condition.     

Field and Coil Design for a Quadrupolar Mirror Hybrid Reactor

A vacuum magnetic field from a superconducting coil set for a single cell minimum B fusion-fission mirror machine reactor is computed. The magnetic field is first optimized for MHD flute stability, ellipticity and field smoothness in a long-thin approximation. Recirculation regions and magnetic expanders are added to the mirror machine without an optimizing procedure. The optimized field is thereafter reproduced by a set of circular and quadrupolar coils. The coils are modelled using filamentary line current distributions. Basic scaling assumptions are implemented for the coil design, with a maximum allowed current density of 1.5 kA/cm². The coils are optimized using a local optimization method and the resulting field is checked for MHD flute stability and maximum ellipticity.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

Stress and Thermal Analysis of the In-Vessel RMP Coils in HL-2M

A set of in-vessel resonant magnetic perturbation(RMP) coils for MHD instability suppression is proposed for the design of a HL-2M tokamak.Each coil is to be fed with a current of up to 5 kA,operated in a frequency range from DC to about 1 kHz.Stainless steel(SS) jacketed mineral insulated cables are proposed for the conductor of the coils.In-vessel coils must withstand large electromagnetic(EM) and thermal loads.The support,insulation and vacuum sealing in a very limited space are crucial issues for engineering design.Hence finite element calculations are performed to verify the design,optimize the support by minimizing stress caused by EM forces on the coil conductors and work out the temperature rise occurring on the coil in diferent working conditions,the corresponding thermal stress caused by the thermal expansion of materials is evaluated to be allowable.The techniques to develop the in-vessel RMP coils,such as support,insulation and cooling,are discussed.     

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.     

Non-inductive Production of ST Plasmas with Washer Gun Sources on the Pegasus Toroidal Experiment

Formation of tokamak-like plasmas via electrostatic helicity injection in the ultra-low aspect ratio Pegasus Toroidal Experiment is reported. Two low-impurity, high-current (1 kA) washer gun current sources have been installed in the lower divertor region. These initially drive current along helical field lines produced by the applied toroidal and vertical fields. At sufficiently low values of externally applied vertical field, the poloidal field generated by the plasma is large enough to cause a poloidal flux reversal. In these cases the plasma relaxes into a tokamak-like configuration. Discharges with I _ϕ≈ 30 kA are produced with less than 2 kA of injected current. These discharges exhibit features indicative of tokamak plasmas, including reversal of poloidal flux at the center column, strong vacuum field deformation, increased current decay times, increased core heating, and characteristic MHD modes common to other helicity-injection-driven toroidal devices.     

Plasma profile and shape optimization for the advanced tokamak power plant, ARIES-AT

An advanced tokamak plasma configuration is developed based on equilibrium, ideal MHD stability, bootstrap current analysis, vertical stability and control, and poloidal field coil analysis. The plasma boundaries used in the analysis are forced to coincide with the 99% flux surface from the free-boundary equilibrium. Using an accurate bootstrap current model and external current drive profiles from ray tracing calculations in combination with optimized pressure profiles, β_N values above 7.0 have been obtained. The minimum current drive requirement is found to lie at a lower β_N of 6.0. The external kink mode is stabilized by a tungsten shell located at 0.33 times the minor radius and a feedback system. Plasma shape optimization has led to an elongation of 2.2 and triangularity of 0.9 at the separatrix. Vertical stability could be achieved by a combination of tungsten shells located at 0.33 times the minor radius and feedback control coils located behind the shield. The poloidal field coils were optimized in location and current, providing a maximum coil current of 8.6 MA. These developments have led to a simultaneous reduction in the power plant major radius and toroidal field from those found in a previous study [S.C. Jardin, C.E. Kessel, C.G. Bathke, D.A. Ehst, T.K. Mau, F. Najmabadi, T.W. Petrie, the ARIES Team, Physics basis for a reversed shear tokamak power plant, Fusion Eng. Design 38 (1997) 27].     

Thyristor crowbar system for the high current power supplies of ASDEX upgrade

The ohmic heating system and the poloidal field coils of ASDEX upgrade are supplied by 15 thyristor converter units with an installed apparent power of 600 MVA. To protect the thyristor converters against dc overvoltage arising from abnormal operations and resulting damages caused by the large energy stored in the AUG magnet coils an overvoltage protection system was required. The paper describes the motivation for—and the design and testing of the thyristor crowbar system representing the thyristor converter overvoltage protection system. It will present the layout, analyse the results of measurements obtained during commissioning, compare them to the calculated (design) values and report on the first experience of operation on the AUG coils improving the safety of the equipment.     

Power dissipation and energy transfer during testing of the ITER toroidal field model coil

The testing of the ITER toroidal field model coil (TFMC) in the background field of the EURATOM-LCT coil took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16 kA in the LCT coil and 80 kA in the TFMC, respectively. The heat load of both coils, including the eddy current losses in the passive structures and the joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He. About 2% of the stored energy was transferred to the cryogenic system after all the safety discharges of both coils together. Most of the energy (about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum vessel. A computer code, based on the full inductance and resistance matrices, has been developed with SIMULINK™. After validation with experimental data the code has been used to perform circuit analysis and to evaluate the power dissipation and energy transferred to the cryogenic plant and to the external power circuits.     

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.     

A 7 T Pulsed Magnetic Field Generator for Magnetized Laser Plasma Experiments

A pulsed magnetic field generator was developed to study the effect of a magnetic field on the evolution of a laser-generated plasma.A 40 kV pulsed power system delivered a fast(~230 ns),55 kA current pulse into a single-turn coil surrounding the laser target,using a capacitor bank of 200 nF,a laser-triggered switch and a low-impedance strip transmission line.A one-dimensional uniform 7 T pulsed magnetic field was created using a Helmholtz coil pair with a 6 mm diameter.The pulsed magnetic field was controlled to take effect synchronously with a nanosecond heating laser beam,a femtosecond probing laser beam and an optical Intensified Charge Coupled Device(ICCD) detector.The preliminary experiments demonstrate bifurcation and focusing of plasma expansion in a transverse magnetic field.     

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.     

Simulation and test results of low-order band harmonic filters to decrease resonance phenomenon in the KSTAR power system

The superconducting magnet power supply which supplies superconducting magnet coil (SC) with the power to generate plasma during a KSTAR experiment for nuclear fusion research is a nonlinear load. Characteristic harmonics are generally produced since it converts AC to DC using 6 or 12 pulsed operation. However, non-characteristic harmonics or inter-harmonics are generated according to output control characteristics. Also, 95% out of the power generated from superconducting magnet coil is reactive power. Therefore, harmonic and reactive power occurring during operation have some bad influences such as voltage drop, voltage distortion and decrease in power factor on the KSTAR power system, and reactive power compensator (RPC) & harmonic filter (HF) system which is able to compensate harmonic and reactive power at the same time was established and has been operated [1]. However, out of non-characteristic harmonics and inter-harmonics caused by output control characteristics of superconducting magnet power supply, the more compensation volume of the RPC & HF system increases, the more voltage distortion with harmonic expansion is caused by harmonics in a low-order band due to the parallel resonance in a low-order band between the KSTAR power system and the RPC & HF system. As a result, it has serious effect on the injection capacity restriction of the RPC & HF system, the unstable operation control of superconducting magnet coil, and the operation of main cooling facilities. This paper presents reasons of the resonance phenomenon of the KSTAR power system and suggests a design plan of additional facilities for stable operation of the KSTAR power system, and proves their effects through the simulation and test results.     

The Test Facility for the EAST Superconducting Magnets

A large facility for testing superconducting magnets has been in operation at the Institute of Plasma Physics of the Chinese Academy of Sciences since the completion of its construction that began in 1999. A helium refrigerator is used to cool the magnets and liquefy helium which can provide 3.8 K-4.5 K, 1.8 bar-5 bar, 20g/s-40g/s supercritical helium for the coils or a 150 L/h liquefying helium capacity. Other major parts include a large vacuum vessel （3.5 m in diameter and 6.1 m in height） with a liquid nitrogen temperature shield, two pairs of current lead, three sets of 14.5 kA-50 kA power supply with a fast dump quench protection circuitry, a data acquisition and control system, a vacuum pumping system, and a gas tightness inspecting devise. The primary goal of the test facility is to test the EAST TF and PF magnets in relation to their electromagnetic, stability, thermal, hydraulic, and mechanical performance. The construction of this facility was completed in 2002, followed by a series of systematic coil testing. By now ten TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a model coil of the PF large coil have been successfully tested in the facility.     

Pulsed high-current discharge in water: adiabatic model of expanding plasma channel and acoustic wave

This paper presents the results of a theoretical and experimental study of the use of a pulsed discharge in water to obtain a strong acoustic wave in a liquid medium. A discharge with a current amplitude of 10 kA, a duration of 400 ns, and an amplitude pulsed power of 280 MW in water at atmospheric pressure created an expanding acoustic wave with an amplitude of more than 100 MPa. To describe the formation of the discharge channel, an isothermal plasma model has been developed, which made it possible to calculate both the expansion dynamics of a high-current channel and the strong acoustic wave generated by it. Our calculations show that the number density of plasma in the channel reaches 10²⁰ cm^–3, while the degree of water vapor ionization is about 10%, and the channel wall extends with a velocity of 500 m s−1. The calculations for the acoustic wave are in good agreement with measurements.     

CICC Joint Development and Test for the Test Facility

The superconducting joint of the NbTi Cable-in -conduit Conductor (CICC) has been developed and tested on the magnet test facility at Institute of Plasma Physics, Chinese Academy of Sciences. The CICC is composed of (2NbTi+lCu)x3x3x(6+ltube) strands each with 0.85 mm in diameter, which has been developed for a central solenoid model coil. The effective length of the joint is about 500 mm. There have been two common fabrication modes, one of them is to integrate the 2 CICC terminals with the copper substrate via lead-soldering, and the other is to mechanically compress the above two parts into an integrated unit. In the current range from 2 kA to 10 kA the joint resistance changes slightly. Up to now, 11 TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a large PF model coil of PF large coil have been completed via the latter joint in the test facility.