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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 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]

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]

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]

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 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]