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.     

Updating the design of the feeder components for the ITER magnet system

The ITER superconducting magnet system generates an average heat load of 23 kW at 4 K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 10–68 kA to 48 individual coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 1.0 MPa and temperature between 4.3 and 4.7 K, is distributed to the coils and structures through 30 separate feeder lines. The feeders also contain the electrical supplies to the coil, helium supply pipes and the instrumentation lines, and are integrated with the current lead transitions to room temperature. The components consist of the in-cryostat feeders, the cryostat feedthroughs and the coil terminal boxes (CTBs). This paper discusses the functional requirements on the feeder system and presents the latest design concept and parameters of the feeder components.     

Introduction to resonant magnetic perturbation coils of the J-TEXT Tokamak

To investigate the interactions between both the static and rotating resonant magnetic perturbations (RMP) and the tokamak plasma, two sets of coils, namely static RMP (SRMP) and dynamic RMP (DRMP), are constructed on the J-TEXT tokamak. SRMP is reconstructed from TEXT-U and mainly produces static m/n = 1/1, 2/1 and 3/1 resonant perturbation field, where m and n are the poloidal and toroidal mode numbers, respectively. DRMP, newly designed and installed inside the vacuum vessel, can generate pure 2/1 RMP. DRMP is also designed to operate in the AC mode and can produce rotating 2/1 RMP which will be used to study the tearing mode control. Due to the effect of the eddy current in the vacuum vessel wall, the amplitudes of the 2/1 component will be attenuated to about 1/3.6 of the DC value when the operation frequency is larger than 500 Hz. However, DRMP can still provide sufficient large rotating 2/1 perturbation for tearing mode related studies.     

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.     

Initial plasma start-up using partial solenoid coils in Versatile Experiment Spherical Torus (VEST)

Initial plasma start-up experiments based on ohmic discharge using partial solenoid coils located at both vertical ends of a center stack have been carried out in Versatile Experiment Spherical Torus (VEST) at Seoul National University. Ohmic discharges with the help of microwave pre-ionization have been performed according to the pre-programed start-up scenario which was experimentally verified by a series of vacuum field measurements using an internal magnetic probe array. A plasma current of around 0.4 kA has been achieved by ohmic discharge using partial solenoid coils, under the toroidal magnetic field of 0.1 T. The vacuum field calculation and fast camera image have revealed that the small plasma current even with significant amount of loop voltage up to 9.7 V is attributed to the imbalance of poloidal field for equilibrium. Modification of the start-up scenario and upgrade of power supplies are proposed to be carried out in order to achieve higher plasma current in the future experiments.     

AC operation of large titanium sublimation pumps in a magnetic field: Results of the test stand for the W7-X neutral beam injectors

A neutral beam injection (NBI) system is being built for the Stellarator experiment Wendelstein 7-X (W7-X) currently under construction at IPP Greifswald. The NBI system consists of two injectors which are essentially a replica of the system present in the Tokamak experiment ASDEX-Upgrade at IPP Garching. A vacuum system with high pumping speed and large capacity is required to ensure proper vacuum conditions in the neutral beam line. For this purpose, large titanium sublimation pumps (TSP) are installed inside the NBI boxes, consisting of 4 m long hanging wires containing Ti and the surrounding condensation walls. The wires are DC ohmically heated up with 142 A to Ti sublimation temperature. A TSP system has been operated since many years in the AUG-NBI system, sublimating Ti in the pauses between the plasma discharges, when no magnetic field is present. However, at W7-X the superconducting coils generate a magnetic field permanently during experimental campaigns, whose stray B field with a maximum of 30 mT, affects the TSPs. Operated with DC, the wires would be deflected against the surrounding panels due to the Lorentz force. A simple possible solution is heating with AC, which reduces the wire deflection amplitude, inducing a risky wire oscillation. The feasibility of the AC operation in an equivalently strong B field such as the stray B field around W7-X has been demonstrated in a test stand for different AC waveforms and frequencies. Several test campaigns have shown no qualitative difference in the pumping properties between AC and DC operation of the TSP and no critical dynamic behaviour of the wires.     

Development of high voltage power supply for the KSTAR 170 GHz ECH and CD system

A 3.6 MW (66 kV/55 A) DC power supply system was developed for the 170 GHz EC H&CD system in KSTAR. The power supply system consists of a cathode power supply (CPS), an anode power supply (APS) and a body power supply (BPS). The cathode power supply is capable of supplying a maximum voltage of ?66 kV and a current of 55 A to the cathode with respect to the collector using pulse step modulation (PSM). The high voltage switching system for the cathode is made by a fast MOS-FET solid-state switch which can turn off the high voltage to the cathode within 3 μs in the occurrence of gyrotron faults. The APS is a voltage divider system consisting of a fixed resistor and zener diode units with the capability of 60 kV stand-off voltage. The anode voltage with respect to the cathode is controlled in a range of 0–60 kV by turning the MOS-FET switches connected in parallel to each zener diode on and off. For high frequency current modulation of the gyrotron, the parallel discharge switch is introduced between the cathode and anode in order to clamp the charged voltage in the stray capacitance. The BPS is a DC power supply with the capability of 50 kV/160 mA. The nominal operation parameter of BPS was 23 kV and 10 mA, respectively, and the voltage output is regulated with a stability of 0.025% of the rated voltage. The series MOS-FET solid-state switch is used for on/off modulation in the body voltage sychronizing with anode voltage. The parallel discharge switch is also introduced between the body and collector for high frequency RF modulation. This paper describes the key features of the high voltage power supply system of the KSTAR 170 GHz gyrotron as well as the test results of the power supply.     

The 70 kA pyrobreaker for ITER magnet back-up protection

The paper presents the results of development and testing of an explosively actuated circuit-breaker (the so-called pyrobreaker) designed and manufactured at the Efremov Institute [1]. In accordance with the ITER specifications this switch will be used for continuous operation with DC currents up to 70 kA and shall be capable, on command, to transfer this current to a resistive load under a voltage up to 10 kV in less than 1 ms.A number of current commutation tests have been carried out on several prototypes [2]. The last experimental campaign has demonstrated reliable operation of the pyrobreaker with 20% safety margin for the interrupted current and 100% margin for the recovery voltage relative to the ITER requirements.Besides, peak current withstand tests have been performed with pulse currents up to 420 kA generated by the unipolar current generator available at the Efremov Institute.     

High temperature superconductors for the ITER magnet system and beyond

The use of high temperature superconductor (HTS) materials in future fusion machines could increase the efficiency drastically, but strong boundary conditions exist. To outline the prospects, challenges and problems, first the benefit of using HTS materials is estimated considering the saving in cryogenic power. Next, it is demonstrated that industrial available HTS materials can be used for fusion today. For this purpose, we give a short summary of results that have been obtained from an ITER conform 70 kA HTS current lead that was designed, built and tested by the Forschungszentrum Karlsruhe and the CRPP Villigen in the frame of the European Fusion Technology Programme and in cooperation with industry. This current lead consists of an HTS part that covered the temperature range from 4.5 to 70 K and a conventional part, making the connection to room temperature. Because the HTS part had no ohmic losses and poor thermal conduction, the refrigerator power necessary for cooling the current lead was reduced drastically. The saving factor could be calculated to be 5.4 at zero current and 3.7 at 68 kA. The current lead could even be operated at 80 kA and with respect to safety criteria of ITER, a complete loss of He flow was simulated showing that the HTS current lead could hold a current of 68 kA for 6 min without active cooling. These results demonstrate that today existing HTS materials can be used in ITER for current leads or bus bar systems.For fusion machines beyond ITER, the development of an HTS fusion conductor would be the key to operate the complete magnet system at higher temperatures. The option of developing fusion conductors based on Bi-2223 and YBCO are briefly discussed. For a success of such conductors, the AC loss optimisation is crucial.     

Fabrication of new joints for SST-1 TF coil winding packs

The Toroidal Field (TF) magnet system of SST-1 has sixteen NbTi/Cu based coils with about one hundred Inter-Pancake (IP) and Inter-Coil (IC) joints. New box type helium leak tight, low DC resistance joints have been designed, fabricated and tested at 5 K temperature and 10 kA DC transport current. The prototype of this joint has been validated in laboratory as well as on spare TF coil winding pack. Moreover, the performance of these joints has been realised and validated on actual sixteen TF winding packs, the joint resistance of ～0.5 nΩ repeatedly measured on hundreds of IP joints. The quality of terminations and joints was ensured at various stages of fabrication. The quality of joint box material was ensured by visual inspection, chemical analysis, radiography test, ultrasonic test, eddy current test, etc. This paper describes joint design drivers, joint design detail, prototype joint fabrication processes, quality assurance (QA)/quality control (QC) adopted during prototype and actual joint fabrication process, joint resistance measurement on actual TF coils and analysis of measured joint resistance in detail.     

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I_p of 611 kA) is 0.01 MW/m² while that in 2011 campaign (with 12 s pulse duration and I_p of 630 kA) is about 0.02 MW/m². The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.     

Bridge joint fabrication and validation for SST-1 PF coil winding pack

A novel concept of bridge joint for Poloidal field (PF) magnet of SST-1 with damaged winding pack has been realized. This joint has been fabricated on 5th and 6th layers of PF#3T coil winding pack (WP) after validation at 10 kA at liquid helium temperature of 4.2 K in current lead test chamber. The joint resistance of bridge joint was measured ∼1.6 nΩ at flat top DC current of 10 kA. This type of joint could be economically useful for revival of a shorted and damaged WP superconducting PF magnets of Tokamaks. In this paper, details of bridge joint design, fabrication and validations are discussed.     

R&D and maintenance on graphite tile of divertor region at EAST

EAST, with full superconducting magnetic coils, had been designed and constructed to address the scientific and engineering issues under steady state operation. The in-vessel components are full graphite tiles as first wall had been operated successfully. In the experiment campaign of 2010, the H mode operation and 1 MA operation have been gotten on EAST. However, in some case, some of the graphite tiles of divertor region are damaged with the plasma parameter enhanced. As most of the damaged graphite tiles are in the divertor region, they are probably damaged by the electro-magnetic force of the halo current when the VDEs occur. The force of the halo current is re-estimated. The structure analysis has been done by the ANSYS software. From the analysis result. It can be obtained that the stress is larger than the allowable stress when the halo current on the graphite tile is larger than 2.7 kA. The tensile testing of the graphite also has been done. As the result, the graphite tiles are damaged when the forces are up to 2400 N. To deal with the problem, two proposes are accepted. In the one hand, the new type graphite material is used, whose tensile strength is up to 45 MPa. In the other hand, the structure of the graphite tiles is optimized.     

Mitigation of edge localised modes with magnetic perturbations in ASDEX Upgrade

The first stage of a significant enhancement of the ASDEX Upgrade experiment with in-vessel coils for non-axisymmetric magnetic perturbations is now operational. First experiments have shown that ELM mitigation can be achieved using various perturbation field configurations with toroidal mode numbers n = 1, 2, 4. The main access criteria is the plasma edge pedestal density to exceed a threshold, which takes the lowest value of about 60% of the Greenwald density for resonant |n| = 1 perturbations. In H-mode plasmas, mode locking or error field-induced magnetic islands are generally not observed. Due to the low local shear of the plasma magnetic field in the vicinity of the perturbation coils around the outboard midplane, the magnetic perturbation is resonant simultaneously on several rational surfaces. It is hypothesised that the existence of image currents on these surfaces ensures good shielding of the error field in the confined plasma.     

ASDEX Upgrade enhancements in view of ITER application

The aim of the ASDEX Upgrade (AUG) programme is to support the design, prepare the physics base and develop regimes beyond the baseline of ITER and for DEMO. Its ITER-like geometry, poloidal field system, versatile heating system and power fluxes make AUG particularly suited.After the transition to fully tungsten coated plasma facing components AUG could be operated without prior boronizations and a low permanent deuterium retention was found qualifying W as wall material. ITER-like baseline H-modes (H₉₈ ～ 1, β_N ～ 2) were routinely achieved up to 1.2 MA plasma currents. W concentrations could be kept at an acceptable level of <5 × 10^?5 by central wave heating (enhancing impurity outward transport) and ELM pacing with gas puffing. The compatibility of high performance improved H-modes, the ITER hybrid scenario, with an un-boronized W wall was demonstrated achieving H₉₈ ～ 1.1 and β_N up to 2.6 at modest triangularities δ ≤ 0.3. This performance is reached despite the gas puffing needed for W influx control. Increasing δ to 0.35 allowed at even higher puff rates still a H₉₈ ～ 1.1.Reliable plasma operation in support of ITER comprised the demonstration of ECRF assisted low voltage plasma start-up and current rise at toroidal electric fields below 0.3 V/m resulting in a ITER compatible range of plasma internal inductance of 0.71–0.97. Disruption mitigation is feasible using strong gas puffs, and the achieved electron densities approach values needed for runaway suppression.Present hardware extensions in support of ITER include the upgrading of ECRH by a 4 MW/10 s system with large deposition variability (tuneable frequency between 105 and 140 GHz, real-time steerable mirrors) for central heating and MHD mode control. A powerful system of 24 in-vessel coils produces error fields up to toroidal mode number n = 4 for ELM suppression and mode rotation control. In connection with a close conducting wall they will open up the road for RWM stabilization in advanced scenarios. For those we are considering LHCD for current drive and profile control with up to 500 kA driven current. The tungsten sources are dominated by sputtering from intrinsic light impurities, and the W influx from the outboard limiters are the main source for the core plasma. ICRH induced electric fields accelerate light impurities, restricting the use of ICRH to just after boronization. 4-strap antennas imbedded in extended wall structures might solve this problem. Finally, doubling the plasma volume with plasma currents above 2 MA in AUG could be the solution for a needed ITER satellite.     

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.     

Performance and modelling of 70kVdc power supply with solid-state crowbar

The experimental activities of tokamak research involve development of high power RF and microwave sources for fusion related heating and current drive applications. High power RF and microwave tubes like Klystron, Gyrotron and Tetrode are in general operated with high voltage DC power supplies. These HVDC power supplies of the order of 70 kVdc, must be equipped with necessary arc fault protection in addition to general over current and over voltage protection.The arc fault protection must act within few microseconds to prevent permanent damage to the RF tube, window, etc. When an arc fault is detected, output voltage of the DC power supply is short circuited using a crowbar device (generally Ignitron, Thyratron, thyristor, rail-gap, etc.) that operates in few microseconds. This diverts the fault current from the load to crowbar device, thereby protecting the load. This is necessary as conventional protection in the power supply input takes ～100 ms to switch-off. The crowbar device must be able to take the fault current till the circuit breaker placed at power supply input is switched off.The arc fault protection is tested for its effectiveness by “wire-burn” test. Full power short circuit of ～1.5 MW DC power supply puts enormous stress on the power supply, utility and the crowbar, therefore frequent wire-burn testing is to be avoided.This report presents simulation of wire-burn test using PSIM software. Optimization of the component values without conducting actual wire-burn test could be achieved.     

AC loss measurements on Cable in Conduit Conductor with random-changing magnetic fields

Beside the generation of the poloidal component of the field, the main function of the poloidal field coils in a tokamak is the control of the shape and the position of the plasma, according to the chosen plasma scenario. A plasma scenario, namely a sequence of plasma shapes, is obtained and controlled by varying the current in the PF coils. The control currents create magnetic fields having complex trends, being almost random waveforms with frequencies in the range between 0.1 and 10 Hz. As a consequence, AC loss is generated in the superconducting coils exposed to those signals, and the feasibility of a plasma scenario is strictly related to the ability to withstand and remove the heat coming from the AC loss.In order to study what the behavior of the loss is in random magnetic fields, namely similar to the control fields, a SULTAN sample is tested under two kinds of random field signals. The first signal is obtained by summing several harmonic frequency components, in the range between 0.2 and 6 Hz, having random amplitude. The second waveform is generated by a random function generator and it has a much broader spectrum of frequencies. The tests are carried out by varying also the maximum amplitude of the signals. The results are here discussed and compared to the results of the single frequency AC loss tests, and a correlation between them is studied.     

Design of the dummy coil for magnet power supply

It is necessary to test it on a dummy coil, before using a magnet power supply (MPS) to energize a Poloidal Field (PF) coil in the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The dummy coil should accept the same large current from the MPS as the PF coil and be within the capability of the utilities located at the KSTAR site. Therefore a coil design based on the characteristics of the MPS and other restrictive conditions needed to be made. There are three requirements to be met in the design: an electrical requirement, a structural requirement, and a water cooling requirement. The electrical requirement was that the coil should have an inductance of 40 mH. For the structural requirement, the material should be non magnetic. The coil support structure and water cooling manifold were made of SUS 304. The water cooling requirement was that there should be sufficient flow rate so that the temperature rise ΔT should not exceed 12 °C for operation at 12.5 kA for 5 min. Square cross-section hollow conductor with dimensions of 38.1 mm × 38.1 mm was used with a 25.4 mm center hole for cooling water. However, as a result of tests, it was found that the electrical and structural requirements were satisfied but that the water cooling was over designed. It is imperative that the verification will be redone for a test with 12.5 kA for 5 min.     

Radiation shielding mass saving for the magnet coils of the VISTA spacecraft

Radiation shielding structure of a design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit has been investigated. Fusion power deposited in the inertial confined fuel pellet debris delivers the rocket propulsion with the help of a magnetic nozzle. The nuclear heating in the super conducting magnet coils determines the radiation shielding mass of the spacecraft. It was possible to achieve considerable mass saving with respect to a recent design work, coupled with higher design limits for coil heating (up to 5 mW/cm³). The neutron and γ-ray penetration into the coils is calculated using the S_N methods with a high angular resolution in (r–z) geometry in S₁₆P₃ approximation by dividing the solid space angle in 160 sectors. Total peak nuclear heat generation density in the coils is calculated as 3.143 mW/cm³ by a fusion power of 17 500 MW. Peak neutron heating density is 1.469 mW/cm³ and peak γ-ray heating density is 1.674 mW/cm³. However, volume averaged heat generation in the coils is much lower, namely 74, 163 and 337 μW/cm³ for neutron, γ-ray and total nuclear heating, respectively. The net mass of the radiation shielding for the magnet coils is 200 tonne by a total mass of 6000 tonne of the space craft.