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The latest superconducting magnets (SM) for fusion are mostly force-cooled, mainly because it allows reliable electrical insulation of the coils using vacuum pressure impregnation (VPI). SM of this type have many leads, feeders and coolant tubes, located in cryostat vacuum, which must sustain high voltages, induced on them by fast current changes. However vacuum loss can spoil their insulation. A few such cases occurred during the T-15 tokamak coils testing, initially having bare leads relying upon vacuum. But its loss generated a coil quench, a protecting current dump at high voltage, followed by breakdown and arc. Even leads insulation by Teflon and fiberglass tape wrap proved to be insufficient. Nevertheless, similar tape wrap insulation of leads and feeders (ILF) was used in EAST, KSTAR, SST-1 and W-7X. So far, seven breakdowns occurred during their coil tests at operating voltage \sim<3 kV. Breakdowns never initiated in the coils, but always on their leads, feeders and sensor lines, indicating that their insulation made by tape wrap were too weak. Instead of ILF improvement some projects undertake Paschen tests. These are planned as the baseline for ITER too. But these tests are valid for the coil with open insulated surface, but are not appropriate for the final tests, when insulation should not be exposed to vacuum. Up to now ILF final tests have been done in all devices at 10–21 kV, but only in good vacuum in spite of the fact that such tests could not guarantee safe operation in case of vacuum loss. We propose to increase ILF strength to the same level, as in the coils, using vacuum-tight grounded stainless steel casings filled up by VPI over magnet leads. This will provide reliable and easily testable solid insulation. Besides, casings would exclude He leaks, providing the second vacuum tight barrier over the ILF. Thus it would increase the magnet reliability and would make it possible to avoid the needs of all single coils test. [ABSTRACT FROM PUBLISHER]

The central solenoid (CS) coil system of the KSTAR device consists of four pairs of coils with up-down symmetry and a set of support structure. The CS coil system should endure various loads including thermal and electromagnetic load during operation. One of the most important roles of support structure is to preserve the structural stability of coil system from vertical forces by applying the pre-compression on the coil stack. The main assembly procedure of the coil stack is as follows: 1) coil subassembly and stacking, 2) structure assembly, 3) preloading process by heating shells in the support structure, and 4) measurement on the final dimension. The special jigs and heating system for the preloading were designed to apply pre-compression on the coil stack during the assembly at room temperature. Moreover, a few tens of thermocouples and strain gauges were installed for measuring temperature and thermal expansion of preloading structures. Since the final assembly of the CS coil was successfully completed with about 7.3 MN of preloading in 2006, the CS coil system has been experienced four thermal cycles from 300 K to 5 K until 2011. This paper describes the main results of applying the pre-compression on the coil stack and mechanical behaviors of preloading structures during thermal cycles. [ABSTRACT FROM PUBLISHER]

The Korea Superconducting Tokamak Advanced Research (KSTAR) device is aimed at advanced tokamak (AT) research. Three years have passed since it achieved its first plasma in 2008. Because it is a superconducting machine and is working toward AT research, it has unique features in terms of the machine engineering and operation. The toroidal field (TF) magnet coils are made of \Nb3\Sn, which provide high TFs up to 3.5 T, and have been fully tested. The poloidal field (PF) magnet coils, consisting of both \Nb3\Sn and NbTi, which have a maximum current of 25 kA in their design, were tested up to 15 kA. A thermal hydraulic analysis is being conducted for PF magnet coil operation. All plasma-facing components (PFCs) are equipped with water cooled graphite tiles and have the capability of being baked up to 350 ^\circ\C. A startup scenario, which considered both the effect of the ferromagnetic material in the cable in conduit conductor jacket of the magnet coils as well as a nonferromagnetic up-down asymmetry in the cryostat structure, was developed and demonstrated its effectiveness by the last two year's reliable operations. Passive stabilizers and in-vessel control coils (IVCCs) are key components to realize AT operation in KSTAR. The segmented IVCC coils were connected to form circular coils for internal vertical control in 2010, and diverted plasmas with high elongation (\kappa \sim 1.8,\ \delta > 0.6) were achieved. A neutral beam injection (NBI) system was developed aiming at 2 MW, 300 s per ion source which meets the long-pulse requirement of KSTAR. An NBI ion source with a power of 1.7 MW at 100 kV has been commissioned for 10 s. Finally, ELMy H-modes were successfully produced with 1.3-MW NBI power at a plasma current of 0.6 MA in the 2010 campaign. The first H-mode discharge (#4200) in KSTAR was achieved one year earlier than officially planned and was done at BT = 2\ \T with Ip = 0.6\ \MA in a well-balanced double null configuration after boronization on the PFC. Successful operations in the early days of KSTAR including H-mode experiments revealed the capability of advanced and steady-state operation which is essential for the international thermonuclear experimental reactor (ITER) and future fusion reactors. [ABSTRACT FROM PUBLISHER]

The AC loss in a large superconducting magnet coil shows a tendency to be changed after assembly. For the ac loss measurement of the Korea Superconducting Tokamak Advanced Research (KSTAR) superconducting coils after assembly, several current scenarios, trapezoidal pulses and a DC offset sinusoidal pulses, were applied to the PF1 Upper (U) and Lower (L) coils during the commissioning. The measurement was done once before and once after the plasma discharge experiments. All coil data were obtained by the tokamak monitoring system and helium distribution system which were designed to measure temperature, pressure, and mass-flow at both inlets and outlets of the coils. The PF1 coils of the cable-in-conduit conductor type were made of Nb3Sn superconducting strands, whose winding scheme is 20 layers with each layer having 9 turns. Each has 10 cooling channels for the heat removal by the supercritical helium at 4.5 K. For the trapezoidal pulse tests, the current was increased up to 4 kA with several different ramp rates and a 2 kA DC offset and 0.5 kA sine waves with different frequencies from 0.1 Hz to 0.3 Hz were applied to the coil. According to the analyses, the AC loss was slightly decreased for the same condition after 700 plasma shots. It was believe that such a result was due to reduced inter-strand resistances which changed the transverse resistance between the inter-strands. The coupling time constant was estimated to be 32 ms in the trapezoidal tests and 13.6 ms for the DC offset sinusoidal pulses. The former is larger than the latter because of the effect of the jacket eddy current loss due to Incoloy 908 which is a ferromagnetic material. Considering the jacket eddy current losses, the coupling time constant was recalculated and the value estimated to be about 13 ms for all current wave forms during first commissioning. [ABSTRACT FROM AUTHOR]

To detect quenches in the Poloidal Field (PF) magnet system is more difficult than the Toroidal Field (TF) magnet system due to excessively high inductive voltages generated by PF pulse-currents and plasma currents. According to reference scenarios being considered so far, the maximum voltage across the PF coils is inductively generated up to about 3.5 kV during the start of plasma (SoP) stage in a very short time period. The voltage measured by compensation of the inductive voltage should be below a certain level which is called as the quench voltage threshold. However, the compensated voltage might be higher than the threshold even with the well-designed compensation schemes. Accordingly, the quench voltage threshold and the quench protection delay time should be properly determined for the quench detection not to take a false action which could cause the fast energy discharge. From the quench simulation using the calculation of hot spot temperature and the resistive voltage growth as a function of time, the proper values of the quench detection parameters of the PF magnet system were derived for the maximum hot temperature rise to be limited within 150 K. [ABSTRACT FROM AUTHOR]

Recent commissioning of two major fully superconducting (SC)-shaped tokamaks, Experimental Advanced Superconducting Tokamak (EAST) and Korean Superconducting Tokamak Advanced Research (KSTAR), represents a significant advance in magnetic fusion research. The key to commissioning success in these complex and unique tokamaks was as follows: 1) use of a robust, flexible plasma control system (PCS) based on the validated DIII-D design; 2) use of the TokSys design and modeling environment, which is tightly coupled with the DIII-D PCS architecture for first-plasma scenario development and plasma diagnosis; and 3) collaborations with experienced internationally recognized teams of tokamak operations and control experts. We provide an overview of the generic modeling environment and plasma control tools developed and validated within the DIII-D experimental program and applied through an international collaborative program to successfully address the unique constraints associated with the startup of these next-generation tokamaks. The unique characteristics of each tokamak and the machine constraints that must be included in device modeling and simulation, such as SC coil current slew rate limits and the presence of nonlinear magnetic materials, are discussed, along with commissioning and initial operational results. Lessons learned from the startup experience in these devices are summarized, with special emphasis on ramifications for International Thermonuclear Experimental Reactor (ITER). [ABSTRACT FROM AUTHOR]

In the Korea Superconducting Tokamak Advanced Research(KSTAR) device, the Toroidal Field (TF) coils were manufactured with Nb3Sn, and the bus lines with NbTi, and all the coils are cooled by supercritical helium. During the first campaign, the TF coils were operated with a steady state current of 15 kA. In case of a quench the conductor is being developed due to an unexpected effect on the superconducting strands, the current should be discharged into an external dump resistor for protection of the magnet. The development or recovery of the quench is related to the protection triggering delay time and the quench voltage. The thermo-hydraulic phenomena were simulated by the GANDALF code and the initial quench propagation condition was found from the minimum quench energy and the initial quench zone. In this paper, we determined the quench voltage level and trigger time by analysis of temperature, voltage, and pressure growth, in the magnet and bus lines, with triggering delays of t = 2 s, 3 s, 4 s, and 5 s, by analysing the effects of a temperature and voltage growth of the magnet and bus lines. [ABSTRACT FROM AUTHOR]

To protect the KSTAR (Korea Superconducting Tokamak Advanced Research) superconducting coils against a quench, the quench detection system based on voltage measurement was fabricated. It used a detect/dump scheme which detects the presence of non-recovering normal zones and activates a dump circuit that transfers most of the stored energy into a dump resistor. Even though it is desirable to discharge the coil energy as fast as possible after quench detection, a time delay between the quench detection and the complete dump circuit actuation will be necessary in order that the quench voltage is to be distinguished from various noises. The quench threshold voltage and the delay time should be set before operation for quench voltage detection so that the maximum hot-spot temperature could be limited to 150 K. This paper describes the hardware techniques to prevent the malfunction of the quench detection due to voltage noises arising from the KSTAR operating scenarios. During the first operation of the KSTAR machine, the inductive voltages and other voltage noises were measured and effectively compensated below voltage thresholds. A quench did not occur and the quench detection system was well operated without any false activation. [ABSTRACT FROM AUTHOR]

KSTAR (Korea Superconducting Tokamak Advanced Research) superconducting magnet consists of a CICC (Cable-In-Conduit Conductor) and is cooled down less than 5 K using supercritical helium. The length of CICC is 610 m for TF coil and maximum about 2,500 m for PF coil respectively, especially the cooling channel is about maximum 300 m in PF coil because of continuous winding scheme. There are many cooling channels in KSTAR coils especially 84 channels in TF magnet system and 100 channels in PF magnet system. Flow imbalance affects cool down of magnet and leads to a difficulty of flow control. The pressure drop between CICC terminals has a close relationship with CICC hydraulic characteristic, helium refrigerator's performance and efficiency. The friction factor which is a representative parameter can be obtained under cryogenic operation condition. We attempt to find out the friction factor of KSTAR CS and PF CICC according to the test result of KSTAR CSMC (Central Solenoid Model Coil) 2nd campaign and compare the results with previous tests. The hydraulic characteristics of KSTAR superconducting magnet system like mass flow distribution, friction factor in conductor, pressure drop and etc during CSMC test and initial commissioning of KSTAR are presented. We can confirm the KSTAR CICC's unique hydraulic behavior in states of cool down and current charging period. Also, we expect that measured data will help to operate KSTAR and be a reference for thermo-hydraulic simulation. [ABSTRACT FROM AUTHOR]

An in-situ measurement of the magnetic field generated by the assembled superconducting magnet coils was held by using precision Hall sensors during the commissioning of the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This was done in order to investigate the magnetic influence of Incoloy 908, which is the jacket material for the cable-in-conduit conductors (CICCs) of the Nb3 Sn coils. After the PF coils were discharged from 1 kA the vertical remanent field at the plasma center was more than 10 G, while the TF coils were not energized. The vertical magnetic field generated by the PF coils had a discrepancy of up to 50 G between measurement and the calculation assuming no magnetic influence of Incoloy 908. Thus, non-negligible ferromagnetism was identified and attributed to Incoloy 908. In contrast, most of the hysteresis observed in the magnetic measurements was eliminated when the Incoloy 908 of the TF windings was saturated by the TF coil charging. [ABSTRACT FROM AUTHOR]

The KSTAR components requiring cryogenic helium coolant for superconducting magnet operations are connected to the helium refrigeration system (HRS) through the helium distribution system (HDS), the final helium distribution station. Twenty eight cryogenic valves including 4 quench protecting valves, many sensors such as temperature sensors, pressure transmitters, and flowmeters are mounted on the system. The HDS has to control the 4.5 K supercritical helium (600 g/s) for 30 superconducting coils and their superconducting bus-lines, 55 K helium (270 g/s) pressurized to 18 bars for the thermal shields and the gravity supports, and a maximum 17.5 g/s of liquid helium for 18 current leads, provided from the HRS which has a 4.5 K equivalent cooling power of 9kW. The helium distribution system consists of a helium distribution box (DB), 5 helium transfer lines, a PLC based helium control system (HCS), etc. The helium control system is connected to the KSTAR supervisory control & interlock system. The detailed status regarding the construction, commissioning during first cool-down, and the instrumentation & control (I&C) system are included in this paper. [ABSTRACT FROM AUTHOR]

To achieve the first plasma of the Korea superconducting tokamak advanced research (KSTAR), the KSTAR superconducting coils were tested in advance. As they should operate in excessively low temperature of 4.5 K and high magnetic field environment of 7.5 T, it is crucial to monitor the cryogenic and the structural behaviors of KSTAR device during the commissioning period including a cool-down. The temperatures of the KSTAR toroidal field (TF) coil and the poloidal field (PF) coils were measured during the entire operating period. The mechanical stresses on the TF and PF structures were continuously monitored to check if they go beyond the limiting value calculated through the simulation. The alignment of the KSTAR device was checked by using displacement sensors. The TF coils were successfully supplied with 15 kA DC current for 8 hours, and the maximum 5 kA/s current variation of the PF coils were tested. For the main experiment, the interlock test of the quench detection system for the KSTAR coils was carried out at reduced currents of 1 kA. From these results the quench protection circuit, and the current-flow of the KSTAR superconducting coils proved to be well performed for the first plasma operation. [ABSTRACT FROM AUTHOR]

The magnet system comprises sixteen toroidal field (TF) coils and fourteen poloidal field (PF) coils for KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak. All the TF coils and 10 PF coils (PF1~ 5 UL) are made of Nb3 Sn strands and 4 PF coils (PF6 ~ 7 UL) are made of NbTi strands. Before operating the superconducting coils, it is critical to verify the operation conditions such as thermo-hydraulic parameters and stability boundaries. In this regards, the TF model coil and CSMC (Central Solenoid Model Coil) have been tested and the operating conditions have been analyzed. Before cool down, parameters such as helium flow through conduits, electrical insulation and coil geometry have been measured at room temperature to confirm the manufacturing soundness. During cool down and current charging, thermo-hydraulic and mechanical parameters by using mechanical sensors have been measured. Also measured was the magnetic field strength in various positions of the coil conductor. Measured data have been analyzed and the results have been compared with the model analysis by simulation code. [ABSTRACT FROM AUTHOR]

The Korea Superconducting Tokamak Advanced Research (KSTAR) Central Solenoid Model Coil (CSMC) was tested at the National Fusion Research Center (NFRC) to verify the design and manufacturing engineering and to ensure reliable operation. The CSMC reached 8.6 T at 20 kA DC operation successfully. We also assessed the AC loss of the CSMC by means of both a sinusoidal wave with a DC offset as well as a triangular pulse. We derived a friction factor correlation of CSMC at room temperature and at cryogenic temperature (4 ∼ 6 K). We also investigated the variation of the friction factor during current charging. The KSTAR PF coil simulation code was validated with inlet and outlet helium temperature, mass flow rate, and pressure drop from the experimental results. The operation temperature margin of the CS1 coil of KSTAR was calculated with the revised KSTAR PF coil simulation coil. [ABSTRACT FROM AUTHOR]

The KSTAR Central Solenoid Model Coils (CSMC), which are in the form of split coils with same dimension, have been tested. The CSMC were successfully charged up to 20 kA and down to zero with different ramp rates. Various pulse waveforms were applied to the CSMC to analyse the AC-loss characteristics of the coils. The measurement method was a gas-flow calorimetry. In this work, two types of waveforms, the DC-biased sinusoidal wave (Idc = 2, 4 kA; ΔIac = 1 kA; ƒ = 0.08 ∼ 0.67 Hz) and a single triangular pulse (Imax = 6 ∼ 10 kA; dI/dt = 0.5 ∼ 2 kA/s), are selected and analysed. From the measured data at the one outlet where helium channel 3 and 4 are merged, the AC loss parameters, related with the hysteresis loss and the coupling loss, are calculated. The B-field strength differs depending on the position within the cooling channels. The spatial field variation and the ramp rate were 0.048∼ 1.9T and 8.1∼ 71 mT/s for the DC-biased sinusoidal waves, while 0.29∼ 4.3 T and 0.024∼ 0.86 T/s for the triangular pulses. The measured AC losses are compared with estimated values and the behavior agrees well. The coupling time constant (nτ) varies with the field strength. In case of the triangular pulses, nτ increases by increasing the field amplitude and the maximum value of nτ is 41 ms with Imax = 10 kA. [ABSTRACT FROM AUTHOR]

The KSTAR (Korea Superconducting Tokamak Advanced Research) superconducting magnet system which consists of 16 TF coils and 14 PF coils adopts a superconducting CICC (Cable-In-Conduit Conductor) type conductor. The KSTAR magnet system uses two different types of CICCs--Nb3Sn cable with Incoloy 908 conduit and NbTi cable with 316LN stainless-steel conduit A continuous CICC jacketing system is developed for the KSTAR CICC fabrication and the jacketing system uses the tube-mill process. It consists of forming, welding, sizing and squaring procedures. The welding condition of CICCs and the fabrication process is described. The repair of the CICC is also discussed. [ABSTRACT FROM AUTHOR]

The KSTAR superconducting magnet system consist,s of 16 TF (Toroidal Field) and 14 PF (Poloidal Field) coils. Both of the TF and PF coil systems use internally-cooled Cable-In-Conduit Conductors (CICC). The PF coil system, which consists of 8 coils in the CS (Central Solenoid) coil system and 6 outer PF coils, provides 17 V-s and sustains the plasma current of 2 MA for 20 seconds inductively. PF 1-5 coils use Nb3Sn CICC in an Incoloy 908 conduit and PF 6-7 coils use NbTi CICC in a modified stainless steel 316LN(316LN+). The fabrication of PF7L/U & PF6L/U coils is completed at present. The engineering issues and fabrication process, which are related with the KSTAR PF 6-7 coil development, are presented and discussed in this paper. TF and PF coils are in the fabrication stage for KSTAR completion in the year 2006. [ABSTRACT FROM AUTHOR]

The superconducting (SC) bus-line system is for the current transmission between SC coils and current leads. In the Korea Superconducting Tokamak Advanced Research (KS TAR) device, twelve pairs of current transmission lines are required for 16 TF coils and 7 pairs of PF coils. The current transmission lines consist of cable-in-conduit conductors (CICCs), which are square-shaped and use a SS316LN jacket surrounding 324 NbTi strands and 162 copper strands. The CICCs are cooled with supercritical helium. The bus-line system is divided into out-of cryostat bus-lines (OCBs), in-cryostat bus-lines (ICBs), and interface terminals. The OCBs are integrated into two vacuum ducts, which have a diameter of 500 mm and total length of about 60 m. There are intermediate joint sections in the OCBs for the flexibility of assembly. The interface terminals at both ends of the OCBs are to give mechanical and electric connection, with consideration of the thermal contraction. The heat loads of the thermal shields in OCBs are about 680 W in total. The ICBs to extend the SC coil joints to the cryostat boundary are supported with some special-shaped structures and G-10 clamps to compensate for thermal contraction and to restrain against lateral motion and rotations. [ABSTRACT FROM AUTHOR]

The Korea Superconducting Tokamak Advanced Research (KSTAR) device requires R&D of 20-30 kA overloaded current leads (OCL) for the superconducting poloidal field (PF) magnet. The PF magnet is under zero current state for a very long time, as it is often the case for superconducting magnets used in nuclear fusion. In such case, a simple approach to reduce the cryogenic load would be to design the current leads to be overloaded. Two main problems in this R&D area are the reduction of the cryogenic heat load and the temperature rise due to the current overloading. In this paper, the current overloading tests and results of the numerical analysis for brass leads are presented and discussed in detail. [ABSTRACT FROM AUTHOR]

The Poloidal Field magnet system of the KSTAR consists of three pairs of coils using the cable-in-conduit conductors cooled by forced super-critical helium. The conductor of PF5 is Nb3Sn superconductor with Incoloy 908 conduit, and that of FF6 and PF7 is NbTi superconductor with SS 316LN. The PF coils are self-supporting with regard to radial loads, and they are arranged symmetrically with respect to the equatorial plane. PFS coil is connected to the TF coil structure at eight points by hinges, while PF6 and PF7 coils are connected at sixteen points by flexible plates to allow relative radial movements. In order to investigate the structural integrity, structural analysis has been conducted including buckling and fatigue analyses. The major design loads are dead weight, assembly loads, thermal load due to cool down, and electromagnetic loads under the critical load conditions. It is found that the PF magnet structures can safely withstand all operating conditions. In addition, AISI 316LN can be substituted for JJ1 as the material of the PF coil structures. [ABSTRACT FROM AUTHOR]

The Korea superconducting tokamak advanced research (KSTAR), which is under construction at the National Fusion R&D Center, Korea Basic Science Institute, Daejeon, Korea, has the mission to develop a steady-state capable advanced superconducting tokamak to establish the scientific and technological bases for a fusion reactor. After an intensive R&D program, substantial progress of the KSTAR tokamak engineering had been made on major tokamak structures, superconducting magnets, in-vessel components, diagnostic system, heating system, and power supplies with industrial manufacturers by May 2002. The engineering design has been elaborated to the extent necessary to allow a realistic assessment of its feasibility, performance, and cost. Since May 2003, the project has been in the phase of procurement. The fabrication of main tokamak structure such as vacuum vessel, cryostat, and supporting structures is well progressed. The manufacturing work of superconducting coils is also proceeding favorably. The tokaniak assembly started in July 2003 after site preparation and assembly jig. The start of commissioning is scheduled for June 2006. This paper describes the key features and engineering progress of the KSTAR tokamak and elaborates the work currently underway. [ABSTRACT FROM AUTHOR]