Lessons learned from twenty-year operation of the Large Helical Device poloidal coils made from cable-in-conduit conductors

The Large Helical Device (LHD) superconducting magnet system consists of two pairs of helical coils and three pairs of poloidal coils. The poloidal coils use cable-in-conduit (CIC) conductors, which have now been adopted in many fusion devices, with forced cooling by supercritical helium. The poloidal coils were first energized with the helical coils on March 27, 1998. Since that time, the coils have experienced 54,600 h of steady cooling, 10,600 h of excitation operation, and nineteen thermal cycles for twenty years. During this period, no superconducting-to-normal transition of the conductors has been observed. The stable operation of the poloidal coils demonstrates that a CIC conductor is suited to large-scale superconducting magnets. The AC loss has remained constant, even though a slight decrease was observed in the early phase of operation. The hydraulic characteristics have been maintained without obstruction over the entire period of steady cooling. The experience gained from twenty years of operation has also provided lessons regarding malfunctions of peripheral equipment.     

Microstructure observations on butt joint composed of Nb3Sn CIC conductors

To precisely evaluate a butt joint technology for the JT-60SA CS coils, microstructure observations on the butt joint composed of Nb₃Sn CIC conductors were conducted using a FE-SEM. As a sample for the observations, the butt joint sample utilized in the joint resistance measurement was used. During the sample fabrication, the butt joint sample was heated up to about 920 K from room temperature for diffusion bonding after heat treatment for Nb₃Sn production. Then, the sample was subjected to the cycles of electromagnetic force in the joint measurement.The observation results indicated that Nb₃Sn strands and a copper sheet were butted properly at the interface of the butt joint. In addition, there were hairline cracks in the Nb₃Sn layers of the strands near the interface. To investigate a cause of the crack initiation, the stresses generated in the butt joint under same conditions were analyzed using a simple model. As a result, the cracks would occur with an axial compressive stress generated by the butt joint fabrication.     

JackPot: A novel model to study the influence of current non-uniformity and cabling patterns in cable-in-conduit conductors

JackPot is a new model that is used to analyse how and to what extend current non-uniformity among strands in a cable-in-conduit conductor (CICC) affects its performance. The joints at the extremities of the CICCs in coils and short samples introduce a non-uniform current distribution among the strands. A detailed and quantitative study down to strand level is required to explain the involved phenomena, to understand their implications on short sample and coil tests and to provide adequate solutions for improvements. The model can be used to evaluate the influence of the joint design and to define its baseline requirements for short-sample qualification testing, and for optimum magnet performance of for example the ITER coils.JackPot is an electrical network model that simulates the interaction between the superconducting strands in the cable (following their precise trajectories), the interstrand contact resistances, the conduit, and the cable’s connection to the joints. The backbone of JackPot is its cable geometry model, from which all relevant properties are derived. All parameters are derived from well defined experimental measurements on conductor sections and joints, except the axial strain for Nb₃Sn strands, which is the only free parameter in the model.The simulations demonstrate that the current non-uniformity is the source for a number of observed phenomena. Another conclusion is that completely filling the bottom joints and upper terminations of a short sample with solder, opposed to only (partly) soldering the cable surface, improves short-sample testing significantly for qualifying the ITER type CICCs. This paper describes the model and gives a few examples of applications for its validation.     

Critical current test results of 13 T-46 kA Nb3Al cable-in-conduit conductor

In the framework of ITER-EDA, a 13 T-46 kA Nb₃Al conductor with stainless steel jacket has been developed in order to demonstrate applicability of an Nb₃Al conductor with react-and-wind technique to ITER-TF coils. Using a 3.5 m sample consisting of a pair of conductors with 0% and 0.4% bending strain, the critical current performances of the Nb₃Al conductors were studied to verify that the conductor achieves the expected performance and the bending strain of 0.4% does not originate degradation. The critical currents were measured at background magnetic fields of 7, 9, 10 and 11 T at temperatures from 6 to 9 K. The expected critical currents were evaluated taking into account the variation of the strain in the cross-section due to the bending strain as well as self-field and non-uniform current distribution as results of an imbalance in the joint resistance and inductances. The calculation results indicate that the current distribution is almost uniform and the experimental results showed good agreement with the expected critical currents. Accordingly, we can conclude that the fabrication process of this conductor is appropriate and the react-and-wind technique using the Nb₃Al conductor is applicable to ITER-TF coils. In addition, the critical current of the Nb₃Al conductor is expected to be 108 kA at 13 T and 4.5 K, resulting in a sufficient margin against the nominal current of 46 kA. Furthermore, it was found that the decrease in the critical current by thermal strain can be made small by applying the bending strain to the conductor so as to reduce the compressive strain at higher fields, i.e. inner side of the coil, in the conductor cross-section.     

DEALS: a demountable superconducting magnet system for fusion reactors

A new type of superconducting magnet system (DEALS) for large fusion reactors is described. Instead of winding large planar or multi-axis coils, as has been previously proposed in fusion reactor designs, the demountable superconducting coils would be made by joining together several prefabricated conductor sections. Conductor types, fabrication processes, and joining schemes are described. The magnet sections would be made at a central factory and shipped to the reactor site for assembly.The magnetic forces on the conductors would be transmitted to an external room temperature support structure via low thermal conductivity bearing blocks. This reduces conductor tensile stresses to very low levels. Differential and mechanical thermal movements between the magnet coil and the external support structure would be accommodated by the use of moveable joints between magnet sections. These pressure type contact joints carry current during magnet operation, and do not carry tensile loads.Finite element analyses on the magnet and its support structure are presented together with analyses of magnet cooling requirements. Results of experiments on small movable pressure type joints at liquid helium temperatures are described.These indicate that adequately low joint losses should be achievable in large magnet systems. Current carrying capcity is not affected by relative motion, and friction coefficients are reaonable. Based on these results and the analyses, the DEALS concept appears feasible for fusion magnet systems.     

Electrical and Superconducting Properties in Lap Joints for YBCO Tapes

The joint process between tapes of coated conductors is a critical issue for most applications of high temperature superconductors (HTSs). In this work several lap joints using different techniques were prepared for three different types of commercially available YBCO-coated conductor tapes, with and without copper stabilizer or stainless steel reinforcement layers. Lap joints with effective lengths in the range of 3 to 20 cm were prepared using low melting point In–Sn and Sn–Pb alloys as soldering materials. The electrical resistance, the critical current, and the n-index of the joints were calculated from the electric field vs. current (E×I) characteristic curves under DC current tests and by further subjecting the same samples to tensile stresses. The results showed that the reinforced tape is the more robust tape for the joint-making process, whereas the copper-stabilized tape presented the lowest joint resistivity but with a relatively smaller mechanical strength against tensile stress.     

ITER CC导体接头测试装置设计与研制

根据国际热核实验反应堆(ITER)校正场线圈(CC)导体接头低温电阻的测试要求,设计并研制了一套用于超导导体接头的低温测试装置。该装置主要包括10 kA超导变压器、低温测试杜瓦、磁体失超保护系统和数据采集系统等。超导变压器的初级线圈及次级线圈采用LHe浸泡的方式进行冷却。超导变压器初级线圈电流引线采用常规铜电流引线,为增加铜的传热面积,采用编织铜引线代替铜棒引线。初级线圈外接磁体电源,利用电磁感应原理,在次级回路感应出超导导体接头测试所需的电流。已经成功进行了一次CC导体接头的低温实验,接头电阻的测试结果分别为8.4纳欧姆和9.3纳欧姆。     

Using the Vincenta code to analyse pressure increases in helium during the quench of a superconducting magnet

The Vincenta code is used to simulate the pressure increases in helium in case of a quench in the superconducting coils. We focus on two classes of coil in which helium is in direct contact with the conductor: coils consisting of cable-in-conduit conductors (as in ITER or JT-60SA), in which supercritical helium is forced through long channels; and bath-cooled coils, in which static helium is confined in short channels perpendicular to the conductor and opening into a bath (as in Tore Supra or Iseult). Various physical phenomena are responsible for the pressure increases in helium, which is subjected to strong heat flux in the conductor during a quench: at the local level, i.e. in the heated channels, the inertial forces that must be overcome to expel the fluid and the friction forces due to the induced velocity; at the global level, i.e. throughout the cryogenic system, the adiabatic compression of non-heated volumes hydraulically connected to the heated channels. Here we analyse the thermohydraulic behaviour of helium to highlight the dominant phenomena, according to the geometry of the helium flow paths. The results are applied to numerical simulation of the pressure rise in case of quench in a JT-60SA cable-in-conduit conductor (CICC) and in the bath-cooled Iseult coil.     

Ten years of cryomagnetic W7-X test facility construction and operation

The construction, commissioning, and operation phases of the W7-X cryomagnetic test facility in CEA Saclay lasted ten years. The large diversity of equipments called, specialties involved and problems solved attest the expertise that was required to operate the test facility and test the coils. Nearly one hundred cryogenic tests were performed on the seventy W7-X coils, at a rate always increasing, using two cryostats each holding two coils.This paper presents the test facility and its operation first, the cryogenic difficulties that were confronted with their solutions, the electro-magnetic difficulties encountered along with corrective actions, and finally the instrumentation and data acquisition aspects.     

A NbTi 10T 388mm bore magnet working at 1.8K in pressurized superfluid helium

A new high field test facility has been constructed for overall property measurements of the conductors developed for the superconducting coils of the TORUS II Tokamak project proposed by the EURATOM-CEA Association. This facility features a dc background field of 10 T in a useful volume of 338 mm in diameter and super-imposed pulsed fields of various configurations.The present paper describes the design, construction and operation of the 10 T field coil system in its pressurized superfluid helium cryostat.     

General atomic's superconducting high field test facility and initial performance

General Atomic has established a high field test facility whose primary mission is to investigate the J-B-T and stability performance margins of commercial NbTi superconductor in the 10 tesla, 4.2 K region. This work is part of the overall DOE/MFE/MAGNETIC SYSTEMS effort to provide an adequate technological base for construction of superconducting toroidal field coils for the next generation of large tokamak fusion devices. The principal components of the facility are the coil/cryostat assembly, the helium refrigerator-liquefier/compressor system, and the gaseous helium recovery and storage system. The epoxy impregnated, layer wound main background field coil generates 8 tesla within its 40 cm diameter bore. The insert background field coil was layer wound with cooling channels provided by "barber pole" mylar conductor insulation. Ten tesla is generated within its 22 cm bore. The initial performance of the facility will be discussed. Future testing calls for operating test coils with implanted heating elements to simulate mechanically induced perturbations. The normal zone growth and recovery behavior will be observed for various disturbance energies. This data will then be compared with results obtained from the transient recovery analysis developed at General Atomic.     

Modelization of the thermal coupling between the ITER TF coil conductor and the structure cooling circuit

The ITER Toroidal Field (TF) coils are required not to quench during the most demanding event: a plasma disruption followed by a fast discharge of the Central Solenoid (CS), the Poloidal Field (PF) coils and the Correction Coils (CC). This event creates large heat deposition in the ITER magnet stainless steel structures in addition to the conductor AC losses. In order to prevent quench occurring in the TF conductor, cooling channels, implemented in the TF coil structure (TFCS), have to remove a large fraction of the heat deposited. The first integrated TF and structure mock-up has been manufactured and then tested in the HELIOS cryogenic test facility (CEA Grenoble) to determine the thermal coupling between the TFCS and the TF conductor, both actively cooled by supercritical helium at 4.4 K and 5 bar. It consists in a stainless steel casing, a cooling pipe glued with resin in the casing groove, winding pack (WP) ground insulation, a radial plate and a copper dummy cable-in-conduit-conductor (CICC). Steady state as well as transient thermal characterizations have been completed in May 2015. Simulation results by thermal hydraulic codes (VENECIA/SuperMagnet) and some of the experimental data are presented and discussed. The thermal coupling between the helium in the cooling tube and the TF coil structure is then modelled as an equivalent heat transfer coefficient in order to simplify the thermal hydraulic (TH) models. Comparison between simplified coupling and detailed coupling is presented.     

The cryogenic system for ITER CC superconducting conductor test facility

This paper describes the cryogenic system of the International Thermonuclear Experimental Reactor (ITER) Correction Coils (CC) test facility, which consists of a 500 W/4.5 K helium refrigerator, a 50 kA superconducting transformer cryostat (STC) and a background field magnet cryostat (BFMC). The 500 W/4.5 K helium refrigerator synchronously produces both the liquid helium (LHe) and supercritical helium (SHe). The background field magnet and the primary coil of the superconducting transformer (PCST) are cooled down by immersing into 4.2 K LHe. The secondary Cable-In-Conduit Conductor (CICC) coil of the superconducting transformer (SCST), superconducting joints and the testing sample of ITER CC are cooled down by forced-flow supercritical helium. During the commissioning experiment, all the superconducting coils were successfully translated into superconducting state. The background field magnet was fully cooled by immersing it into 4.2 K LHe and generated a maximal background magnetic field of 6.96 T; the temperature of transformer coils and current leads was reduced to 4.3 K; the inlet temperature of SHe loop was 5.6 K, which can meet the cooling requirements of CIC-Conductor and joint boxes. It is noted that a novel heat cut-off device for High Temperature Superconducting (HTS) binary current leads was introduced to reduce the heat losses of transformer cryostat.     

Empirical investigation of factors affecting the stability of cable-in-conduit superconductors

Experiments investigating the stability of cable-in-conduit superconductors were performed using test conductors composed of small, quadruplex cables enclosed in steel tubes. To examine the effects of a variation in stabilizer resistivity, all test conductors contained two strands of identical NbTi/Cu composite superconductor and two strands of solid copper. The grade of copper was varied for each test conductor so that a different net stabilizer resistivity existed for each conductor. When these conductors were operated outside the range of multiple stability, tests showed that their stability was negligibly affected by stabilizer resistivity; conversely, the range of operating currents for which multiple stability occurs was affected. However, the surprising result of this study is that stability is apparently affected when a large fraction of the copper stabilizer is segregated into separate cable strands. These observations are compared with other data in the literature.     

A 3D computer code for steady-state magnetic field calculation inair

A 3-D computer code for the calculation of magnetic fields, self- and mutual-inductance coefficients, and electromagnetic forces is considered. A unique feature of this code is the presence of a number of very precise models for the calculation of the field produced by circular or noncircular coils, busbars, or any other kind of conductor of arbitrary shape. The possibility of using different coordinate systems for input/output data represents an additional user-oriented facility. The problems of singularities both for the elliptic integral calculation, used for circular coils, and for Biot-Savart's integral have been taken into account for the field calculation inside the conductors. Since this code has been conceived to solve magnetostatic problems in air-core devices for fusion research, an easy input for particular windings, such as toroidal and poloidal windings, is also provided. A means for automatic generation of the points of the current lines in busbars and in noncircular coils has been inserted in the code as well     

The propagation of the resistive region in high current density superconducting coils

A method is given describing the propagation phenomenon of the resistive region within the winding of superconducting coils wound from semi-stabilized filamentary conductors. The method permits the approximate calculation of the velocities of propagation of the normal conducting front in superconducting coils of high current density. The calculation is carried out assuming that the parameters of the conductor materials, the short sample data, and the heat transfer coefficient from the conductor into the ambient medium are given. The propagation velocities of the normal front have been measured in insulated wires freely immersed in liquid helium, in test samples using conventional interleaving material, and in a compact multilayer coil wound from the same wire. The calculated results are compared with experimental data.     

Dc power system for mutual coupled large scale superconducting coils

The large helical device (LHD) is a fusion experimental facility which has a large scale mutually coupled superconducting coil system and it is now under operation at the National Institute for Fusion Science (NIFS). This paper shows the design and the operation results of the dc power systems to drive these coils. First, the outline of the LHD and the power supplies are described. Next, a control system and the current controller for the power supplies are introduced. Finally, some experimental results are introduced.     

Generalized stiffness coefficients for ITER superconducting cables, direct FE modeling and initial configuration

Superconducting coils are one of the key technical solutions used for generation of high magnetic field in modern tokamaks. Nb₃Sn superconductivity depends not only on temperature and magnetic field as e.g. NbTi, but also on the strain state of the strands inside the conductor. It is hence very important to be able to predict the mechanical deformations due to manufacturing processes and operating conditions. The conductors for ITER, the International Thermonuclear Experimental Reactor currently under construction, have a complex structure that makes analytical estimations of stiffness applicable only for the first cabling stages. In this work, a wide range of numerical simulations has been performed, by using several types of finite element models. This paper shows some analytical estimations for stretching and twisting and compares them with the numerical results of the different models. Some comparisons with experimental tests are also presented. Furthermore, it is shown that direct finite element analyses are compulsory for higher cable stages, but need the knowledge of the initial configuration as precise as possible for meaningful simulations. This problem is also addressed in this paper.     

Evaluation of thermal gradients and thermosiphon in dual channel cable-in-conduit conductors

In an effort to optimize superconductor cryogenics of large coils, dual channel cable-in-conduit conductors (CICC) have been designed. The qualitative and economic rationale of the conductor central channel is here justified but brings high complexity to the conductor cooling characteristics. Temperature gradients in the cable must be quantified to guarantee conductor temperature margin during coil operation under heat disturbance and set adequate inlet temperature. A simple one-dimensional thermal model, with neither fluid nor strand or jacket conduction, allows to better understand and quantify the steady state behavior of CICC central and annular channels. This thermohydraulic model with homogeneous central and annular temperatures and no jacket conduction is summarized with explicit thermal coupling equations. Local convection coefficients chosen proportional to friction factors lead to a model of global interchannel heat exchange coefficient serving the bithermal model. A first stationary experimental evaluation of the internal heat transfer coefficient using the interchannel heat exchange space constant at various heat loads and mass flow rates is illustrated on two full size samples tested at cryogenic temperatures. Annular heaters experiments with low distributed power achieve pertinent model correlation. Discrepancy between model and experimental data may be linked to the simplistic homogeneous annular temperature hypothesis, to the estimate of CICC mass flow distribution among channels, and to gravitational effects at high heat loads. Perturbation due to the thermosiphon generated between the two channels is considered since neither the experiments nor the expected applications are free of gravity.     

Electrical properties of cold-pressing welded NbTi persistent joints

The cold-pressing welding method is employed to fabricate persistent joints between NbTi multifilamentary conductors, and a series of persistent joints are thus made at different mechanical pressures. The electrical properties of these persistent joints are tested by a joint resistance measurement device based on the current decay measurement method. Test results show that the cold-pressing welding method is reliable and applicable to NMR and MRI applications. Experiment results also shows that the joint resistances and current-carrying capacities of the joints seem to have no apparent relevance to the mechanical pressure in a wide range of 4–20 MPa, but the current decay ratio due to flux creep has obvious relevance to the mechanical pressure. Further research is needed to advance this joint manufacturing technology. Besides, measured current increase under 1 T background magnetic field is observed and explained, which has a reference meaning for testing persistent joint resistances.