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.     

Operational experience with forced cooled superconducting magnets

Force-cooled concept has been chosen for ITER superconducting magnet to get reliable coil insulation using vacuum-pressure impregnation (VPI) technology. However 17 breakdowns occurred during operation of six magnets of this type or their single coil tests at operating voltage < 3 kV, while ITER needs 12 kV. All the breakdowns started on electric, cryogenic and diagnostic communications (ECDCs) by the high voltage induced at fast current variations in magnets concurrently with vacuum deterioration, but never on the coils, though sometimes the latter were damaged too. It suggests that simple wrap insulation currently employed on ECDCs and planned to be used in ITER is unacceptable. Upgrade of the ECDC insulation to the same level as on the coils is evidently needed. This could be done by covering each one from ECDCs with vacuum-tight grounded stainless steel casings filled up with solid insulator using VPI-technology. Such an insulation will be insensitive to in-cryostat conditions, excluding helium leaks and considerably simplifying the tests thus allowing saving time and cost. However it is not accepted in ITER design yet. So guarantee of breakdown prevention is not available.     

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.     

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.     

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.     

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.     

Effect of an additional ferromagnetic material on the toroidal magnetic field ripple in the ITER

In the ITER tokamak, the toroidal magnetic field (TF) ripple is estimated with TF coils only, with the installation of ferromagnetic inserts (FIs), and with test blanket modules (TBMs) by using a 2-D code for easy and fast calculation. We assessed the effects of the thickness of the FIs on the TF ripple in order to optimize the FI. And we analyzed how the TBMs distort the TF, and calculated the TF ripple for various amounts of a ferromagnetic material and the positions of the TBMs. Even in the case of moving the TBMs outward up to 60-cm, and reducing the ferromagnetic material to 52%, the TF ripple is not decreased below 0.38%. So we had to adopt ripple correction coils. With a 52% reduced amount of the ferromagnetic material in a TBM, we could reduce the TF ripple to 0.28% at a coil current of 100 kA turn per each coil. And with an outward recess of the TBM up to 60 cm, we could reduce the TF ripple to 0.23% at a coil current of 250 kA turn per each coil. As a combined approach, if we reduce the amount of a ferromagnetic material in a TBM to 30%, and recess the TBM to 15 cm, we can efficiently obtain the TF ripple of 0.25% at a coil current of 150 kA turn per each coil.     

The performance test and analysis of ITER Main and Correction Busbar conductor

The first ITER Main Busbar (MBCN1) and Correction Busbar (CBCN1) conductor samples were manufactured in ASIPP and tested in the SULTAN facility. This paper introduces the sample manufacture, including strand, cabling, jacketing and sample preparation, and discusses the performance of MBCN1 and CBCN1 conductors. The testing results show that both samples have high T_cs, and meet the ITER requirement.Due to the ITER acceptance standard T_cs of MB conductor was changed to 6.7 K at 45.5 kA/3.9 T. The performance of MBCN1 conductor after cyclic load fits the ITER requirement, but the sample was only tested at 57 kA/2.75 T before cycling test. Using some hypothesis and equation to extrapolate the T_cs performance of MBCN1 conductor before cycling test, the result also fits the ITER requirement.For CBCN1 conductor, the central line of the central cooling spiral shifted about 1.3 mm during the cabling. The deviation causes an increase of the max self-field by about 0.005 T, which could not influence the CBCN1 conductor real T_cs performance at peak field.     

Mechanical test on the ITER CC and Feeder jacket

This paper focuses on mechanical tests on the ITER correction coils (CC) and Feeder jacket 316L stainless steel material. During manufacture, the conductor will be compacted and spooled after cable insertion. Therefore, sample jackets were prepared under compaction in order to simulate the status of conductor during manufacturing. Yield strength (0.2% offset), ultimate tensile strength, Young's modulus and elongation at failure shall be reported. The mechanical properties of materials were measured at 300 K and low temperature (<7 K). The cryogenic test results show that the present jackets have very high properties. It is concluded that the results meet the ITER requirement.     

Design of the power system for dynamic resonant magnetic perturbation coils on the J-TEXT tokamak

A set of in-vessel saddle coils has been installed on J-TEXT tokamak. They are proposed for further researches on controlling tearing modes and driving plasma rotation by static and dynamic resonant magnetic perturbations (RMPs). The saddle coils will be energized by DC with the amplitude up to 10 kA, or AC with maximum amplitude up to 5 kA within the frequency range of 1–5 kHz. At DC mode two antiparallel 6-pulse phase thyristor rectifiers are chosen to obtain bidirectional current, while at AC mode an AC–DC–AC converter including a series resonant inverter can generate current of various amplitudes and frequencies. The paper presents the design of the power supply system, based on the definition of the power supply requirements and the feasibility of implementation of the topology and control strategy. Some simulation and experimental results are given in the end.     

Designing and manufacturing of the coil support structure of W7-X

Wendelstein 7-X (W7-X) is a fully optimized low-shear stellarator and shall demonstrate the reactor potential of this fusion plant. It is presently under construction at the Greifswald Branch Institute of IPP. The superconducting magnet system will allow continuous operation, limited only by the plasma exhaust system whose capacity is designed for 30 min full power operation. The Wendelstein 7-X (W7-X) coils and structures are part of the largest superconducting fusion device being constructed at present. They represent a technical challenge at industrial level and the need for proven techniques and manufacturing processes in accordance to the highest quality standards. The production of these components requires a management of monitoring for quality and tests. The coil system consists of 20 planar and 50 non-planar coils. They are supported by a pentagonal 10 m diameter, 2.5 m high coil support structure (CSS). The CSS is divided into five modules. Each module consists of two equal half modules. The manufacturing status of the CSS and the main project management and technical challenges will be presented. The lessons learned in the large scale production of this difficult kind of support structure will be presented as relevant experience for the realization of similar systems for future fusion devices, such as ITER.     

Revisiting the poloidal magnetic confinement

This note proposes a closed poloidal magnetic configuration with an in-vessel coil system held by shielded supports. A dipole field is bounded by external coils and constrained into a hollow torus aiming at uniform intensity. In the horizontal mid-plane region the external coils and the dipole outer coils are broken in four arcs and bridged by couple of straight branches. Arcs and straight branches build a set of four side coils. In the clearance between their straight branches four tunnels in the poloidal magnetic field are achieved, to pass the supports and the feeders of the in-vessel coil system.A poloidal machine with a plasma thick as those of present large experiments is outlined. The dipole radius is 5.4 m, the plasma about it has a constant poloidal cross-section about 40 m², a volume about 1300 m³ and a minimum thickness 1 m in the outboard. The magnetic field ranges from 1.4 to 1.8 T.     

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.     

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.     

Completion of designing and manufacturing of the coil support structure of W7-X

In February 2000, the project called coil support structure for the Wendelstein 7-X fusion machine was started. Since October 2009 the full production of this big (80 tons) and complex component is now completed and delivered at IPP Greifswald. The W7-X coil system consists of 20 planar and 50 non-planar coils. They are supported by a pentagonal 10 m diameter, 2.5 m high called coil support structure (CSS). The CSS is divided into five modules and each module consists of two equal half modules around the radial axis. Currently, the five modules were successfully assembled with the coils meeting the tight manufacturing tolerances. Designing, structural calculation, raw material procurement, welding & soldering technologies, milling, drilling, accurate machining, helium cooling pipe forming, laser metrology, ultra sonic cleaning and vacuum test are some of the key points used all along this successful manufacturing process. The lessons learned in the large scale production of this difficult kind of support structure will be presented as relevant experience for the realization of similar systems for future fusion devices, such as ITER.     

ITER operational space for full plasma current H-mode operation

Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I_p^{^}4. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I_p of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.     

Fatigue assessment of the ITER TF coil case based on JJ1 fatigue tests

The material of the TF coil case in the ITER requires to withstand cyclic electromagnetic forces applied up to 3 × 10⁴ cycles at 4.2 K. A cryogenic stainless steel, JJ1, is used in high stress region of TF coil case. The fatigue characteristics (S–N curve) of JJ1 base metal and welded joint at 4.2 K has been measured. The fatigue strength of base metal and welded joint at 3 × 10⁴ cycles are measured as 1032 and 848 MPa, respectively. The design S–N curve is derived from the measured data taking account of the safety factor of 20 for cycle-to-failure and 2 for fatigue strength, and it indicates that an equivalent alternating stress of the case should be kept less than 516 MPa for the base metal and 424 MPa for the welded joint at 3 × 10⁴ cycles. It is demonstrated that the TF coil case has enough margins for the cyclic operation. It is also shown the welded joint should be located in low cyclic stress region because a residual stress affects the fatigue life.     

Commissioning and initial operation of KSTAR superconducting tokamak

The commissioning and the initial operation for the first plasma in the KSTAR device have been accomplished successfully without any severe failure preventing the device operation and plasma experiments. The commissioning is classified into four steps: vacuum commissioning, cryogenic cool-down commissioning, magnet system commissioning, and plasma discharge.Vacuum commissioning commenced after completion of the tokamak and basic ancillary systems construction. Base pressure of the vacuum vessel was about 3 × 10^?6 Pa and that of the cryostat about 2.7 × 10^?4 Pa, and both levels meet the KSTAR requirements to start the cool-down operation. All the SC magnets were cooled down by a 9 kW rated cryogenic helium facility and reached the base temperature of 4.5 K in a month. The performance test of the superconducting magnet showed that the joint resistances were below 3 nΩ and the resistance to ground after cool-down was over 1 GΩ. An ac loss test of each PF coil made by applying a dc biased sinusoidal current showed that the coupling loss was within the KSTAR requirement with the coupling loss time constant less than 35 ms for both Nb₃Sn and NbTi magnets. All the superconducting magnets operated in stable without quench for long-time dc operation and with synchronized pulse operation by the plasma control system (PCS). By using an 84 GHz ECH system, second harmonic ECH assisted plasma discharges were produced successfully with loop voltage of less than 3 V. By the real-time feedback control, operation of 100 kA plasma current with pulse length up to 865 ms was achieved, which also meet the first plasma target of 100 kA and 100 ms. The KSTAR device will be operated to meet the missions of steady-state and high-beta achievement by system upgrades and collaborative researches.     

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.     

Neutronics analysis and nuclear heating measurement up to the TFC in a mock-up of the ITER inboard shield

A mock-up of ITER including the inboard shield, the vacuum vessel and the coil region, was set up at ENEA Frascati and irradiated with 14 MeV neutrons produced by the Frascati Neutron Generator (FNG). The mock-up dimensions and materials composition are consistent with the current ITER design.The primary objective of the experiment was to validate the MCNP calculations (C) of nuclear heating measured (E) in the region corresponding to the ITER Toroidal Field (TF) coil. An accuracy on C/E ratio ≤±10% was required.The neutronics and shielding properties of the mock-up were also studied throughout the measurement of selected activation reaction rates up to about 1 m depth. Due to the very low activity induced in the foils, the measurements in the deepest experimental positions were performed at the underground low background facility of the Laboratori Nazionali del Gran Sasso, using ultra-low background high purity germanium (HPGe) detectors.The measured reaction rates and nuclear heating were thus compared with the results of the Monte Carlo code MCNP5 coupled with the FENDL-2.1 library.