Cable-in-conduit Conductor Research Articles

DC performance and AC loss of sub-size MgB2 CICC conductor for fusion magnet application

Given the low price and relatively high transition temperature (39 K) of MgB2 conductor, MgB2-based superconductors are a potential candidate for the lower field fusion coils, such as poloidal field (PF) coils, correction coils and feeders. However, to date, the application of MgB2 is limited to demonstrators in a low magnetic field of up to 5 T and at temperatures of up to 10 to 20 K, relying on cryogen-free, helium gas or liquid hydrogen cooling, which significantly reduce the cost of cryogenic systems. To demonstrate the feasibility and performance verification of large size MgB2 PF conductors based on ITER and CFETR requirements, a 4th-stage subsize MgB2 cable-in-conduit conductor (CICC) cable sample is made at the Institute of Plasma Physics, Chinese Academy of Sciences. The CICC contains 96 in situ MgB2 superconducting wires, manufactured by Western Superconducting Technology Ltd. and 48 copper wires. The critical current of the sub-size cables and MgB2 witness wires are examined with different background magnetic fields at 4.2 K. In addition, the AC loss is measured utilizing magnetization and calorimetric methods. To further clarify the influence of electromagnetic force on the AC loss performance, the cable sample is pressed transversely at room temperature and then inserted into a dipole magnet for AC loss measurement at 4.2 K. The critical current at 4.2 K of the subsize MgB2 CICC cable shows 20% degradation compared to the witness wires at 2 T background magnetic field. However, no further critical current degradation is visible during ramping up and down the magnetic field. The coupling loss time constant for 1 T background magnetic field amounts to 480 ms. No significant effect of the applied transverse stress on the coupling loss is observed between 0 and 10 MPa.

High-energy synchrotron X-ray tomography coupled with digital image correlation highlights likely failure points inside ITER toroidal field conductors

Two sections of heat-treated (HT) and non-heat-treated (NHT) Cable-in-Conduit Conductor (CICC) of a design similar to the ITER tokomak have been imaged using very high energy X-ray tomography at the ESRF beamline ID19. The sample images were collected at four temperatures down to 77 K. These results showed a greater degree of movement, bundle distortion and touching strands in the NHT sample. The HT sample showed non-linear movements with temperature especially close to 77 K; increasing non-circularity of the superconducting fibre bundles towards the periphery of the CICC, and touching bundles throughout the CICC. The images have highlighted where future design might improve potential weakness, in particular at the outer perimeters of the conductor and the individual sub-cable, ‘petal’ wraps.

Performance of first Bi-2212 cable with pre-over pressure and over pressure heat treatment

Due to the excellent current carrying performance of Bi2Sr2CaCu2O8+x (Bi-2212) and the development of its industrial manufacturing technology, Bi-2212 is a promising material to be developed as superconductor for application in fusion reactor magnets. The cable-in-conduit conductor (CICC) concept is often chosen for the development of large-scale magnets because of their high stability. Bi-2212 is presently the only kind of copper oxide superconducting material which can be made into solid round wire, which provides a good basis for developing CICCs. The over pressure (OP) heat treatment can significantly improve the superconducting performance of Bi-2212 wires but it also reduces the wire diameter by ∼5%. This leads to an increase of the void fraction of CICCs, typically from 30% to 40% for a CICC with ITER scale dimensions. A pre-OP heat treatment before OP is proposed in this study. The reduction of the wire diameter can be completed before the formation of the continuous superconducting phase, which would dramatically decrease the CICC void fraction. One Bi-2212 cable consisting of 84 wires, was first pre-OP heat treated successfully and after completing the OP heat treatment, the cable’s transport performance was tested. The results showed good performance with a critical current (I c) of 35.7 kA at 5.8 T background field in 4.2 K, which is consistent with the predication.

Qualitative Study On Cable Breakage of Nb3Sn CICC Based On Direct Current Potential Drop Method

Nb3Sn cable-in-conduit conductor (CICC) technology is widely applied to the large-scale superconducting magnets. Due to the multi-level composite structure, it is easy to cause partial damage or even strand breakage of the superconducting cable in the complex deformation mode. To evaluate the cable breakage of Nb3Sn CICC, direct current potential drop (DCPD) technique is proposed in this study. Serious of benchmark samples of cable with stage 3 were prepared and different breakage rates were designed in the cable samples. Potential difference (PD) obtained from the both side of the cable breakage was measured and analyzed. The experimental results indicated that direct current potential drop (DCPD) method is effective to qualitative measurement of cable breakage.

Cooling Loop Design of Background Magnet of 15T Conductor Performance Test Facility

“China Fusion Engineering Test Reactor (CFETR)” is a next-generation tokamak in China, and superconducting magnets are the cornerstone of the CFETR. Manufacturing its superconducting magnets requires a higher performance Cable-in-Conduit Conductor (CICC), the conductor performance test facility is designed to solve this problem. The purpose of the conductor performance test facility is to develop and test high-performance CICCs. The background magnetic field of the facility is provided by Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn superconducting magnets, and it is a big challenge to build this background magnet since its maximum magnetic field is 16.5T. The heat load of the background magnet and the friction factor of CICC are analyzed and calculated in this paper, the reasonable parameters of the cooling channels were determined and provided the basic variables for the subsequent engineering design and optimization.

Fabrication and Critical Current Evaluation for HTS Twisted Cables Consisting of Soldered-Stacked-Square (3S) Wires

High temperature superconducting (HTS) conductors or cables can be considered as a significant candidate in various large-current HTS applications. To develop large-current HTS conductors or cables, HTS wires of small square dimension have been suggested and studied in our group, and the wire is named Soldered-Stacked-Square (3S) wires. Based on 3S (2+10c)-wires composed of two HTS tapes and ten copper tapes, a novel HTS twisted cable structure named “2mm-7-cable” including seven (2+10c)-wires was firstly proposed in this study. This structure is most similar to the cable in conduit conductor (CICC) structure of low temperature superconductor (LTS). To begin with, we introduced the fabrication process of the 2mm-7-cables, including preparation of soldered-stacked-square (3S) wire, winding copper wire process, twisting and cabling process. Then, their critical current performance was evaluated experimentally. The test results of critical current show that when the voltage-lead is directly connected to the surface of the (2+10c)-wire after removing the copper wires of both ends, the measured resistance of 2mm-7-cable will be greatly reduced. And the summation of the critical currents of seven (2+10c)-wires are almost identical to the one of the 2mm-7-cable. Thus, the feasibility of the fabrication method for the novel HTS twisted cable is partially verified in this study.

Preliminary Design of DC Magnet for Super-X Test Facility

The engineering design of the China Fusion Engineering Test Reactor (CFETR) has been completed in 2020. A largescale R&D project for CFETR named the comprehensive research facility for fusion technology (CRAFT) was initiated in Sep. 2019. As a sub-task of the CRAFT, a superconducting experiment testing platform (Super-X for short) is planned to be constructed for evaluating the future superconducting parts performance. Its main technical parameter includes: the background field of 15 T, the dimensions of testing space of 100 × 160 × 550 mm, the field homogeneity of testing space ≥ 95%, and the maximum testing current of 100 kA. For now, the preliminary design for Super-X has been accomplished. The DC magnet provides the background magnetic field for Super-X. It is a kind of split coil that includes two symmetrical coils assemblies. The innermost diameter is about 644 mm, the outermost diameter is 2874 mm. Each coil assembly is composed of three concentric Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn coils. The peak field is 15.7 T and the field homogeneity reach to 98.2%. The high field coil (HFC) and medium field coil (MFC) is typically dominated by high-J <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn cable-in-conduit conductor (CICC), and the low field coil (LFC) by ITER grade Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn CICC. This paper presents the preliminary design of the background DC magnet system (BM).

Transverse Stress Applying Technology for High-Current Superconducting Conductors Testing

High-current superconducting conductors are widely used in fusion magnets, high energy accelerators and high magnetic field facilities. It's urgent to carry out experimental research on DC performance of new conductors to meet the development needs of these magnets. A new test facility with maximum background magnetic field of 6 T and maximum sample current of 30 kA is developed. For the purpose of preventing the sample conductors from premature quenches due to mechanical disturbance under the magnetic field, it is necessary to apply a certain transverse stress to the sample through sample support. In this paper, two different kinds of transverse stress applying technologies, active method and passive method, are developed. The general design of these two kinds of sample supports, the influence of cooling and powering process on sample transverse stress with active sample support, the assembly gap analysis of passive sample support are described in detail. The used sample supports respectively for Nb-Ti Rutherford cable, MgB <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sub> Cable-in-conduit Conductors (CICC) and Bi-2212 sub-cable testing are presented. Comprehensively considering the stability margin of conductor sample and the influence of transverse stress on performance of the sample, the selection basis of transverse stress applying at room temperature is also summarized.

An analytical model for coupling losses in large conductors for magnetic fusion

During the operation scenarios of large superconducting tokamaks such as JT-60SA or ITER, disruptions or VDE (Vertical Displacement Events), AC losses are deposited in the superconducting magnets. Due to these AC losses, the temperatures of the Cable in Conduit Conductors (CICC) increase, reducing the temperature margins. It is of highest importance for the operation to correctly estimate AC losses and conductor temperatures. Due to the fast transient nature of those events, an important component of the losses is the coupling losses.The approach for coupling losses estimation, using a unique effective time constant “nτ” is not completely adequate to calculate these losses.For this purpose, since 2010, a heuristic analytical model called MPAS (Multizone PArtial Shielding), has been proposed. It is now currently used by several teams.For each conductor of the coils of a tokamak, the parameters of the MPAS model and nτ are estimated based on the characterization of the conductor AC losses under a sinusoidal excitation at different frequencies. This process is extensively conducted in existing facilities such as the SULTAN test facility at SPC (Swiss Plasma Center), on the dipole and press of Twente University or on JOSEFA dipole test facility at IRFM (Research Institute for Magnetic Confinement) at Cadarache.This characterisation leads to the constitution of a database for the different conductors used in the tokamak. This has been pushed to the uttermost within the ITER qualification and QA/QC (Quality Assurance, Quality Control).The methodology for estimating the MPAS parameters in relation with the measured AC losses is described in details and illustrated for two conductors. Using this database, it is then possible to calculate the coupling losses during ITER module discharges and scenarios. Estimation of coupling losses are compared to experimental results in recent experiments on ITER CS (Central Solenoid) modules.

Critical bending wavelengths associated with current sharing temperature degradation in multistage twisted Nb3Sn cable CICC

The concept of a bending wavelength has been proposed in the description of strand deformation phenomena arising from electromagnetic loads, in cable-in-conduit conductor (CICC) with multistage twisted Nb3Sn cable, that result in the degradation of the CICC current sharing temperature (Tcs). A survey of CICC Tcs performance covering Nb3Sn CICC samples from the recent past two decades shows interesting results that can help relate bending wavelength to CICC parameters such as the void fraction (Vf) and the twist pitches of the CICC cable to which the latter has attributed to Tcs degradation according to prior research. In this paper, a simple cable strand model is used to link CICC parameters especially the CICC Vf to the bending wavelength of the CICC strand. Through this treatment, a range of bending wavelengths are identified that can be associated to characteristically different Tcs performance behaviour of the CICCs. The bending wavelength values corresponding to transitions between these characteristically different CICC Tcs behaviours are identified as critical bending wavelengths. Additionally, a picture of the internal mechanism resulting in Tcs change is presented along with comparisons of bending wavelength values derived from different theories and the value from destructive examination identified in the literature for Nb3Sn CICC samples.

A FRACTAL PRESSURE DROP MODEL FOR CABLE-IN-CONDUIT CONDUCTORS BASED ON POROUS MEDIA ANALOGY

Assessment of the pressure drop that occurs in cable-in-conduit conductors (CICCs) requires a reliable predictive model. In this paper, a fractal model to predict the pressure drop of helium flows in CICCs is proposed based on the porous media analogy and fractal theory. A fractal friction factor correlation is also derived using the proposed pressure drop model. The good agreement between the predictions and the experimental data for conductors with [Formula: see text] verifies the validity of both the fractal pressure drop model and the friction factor correlation for such configuration conductors. The fractal correlation indicates that the friction factor of the cable bundle in a CICC is related not only to the void fraction but also to the average hydraulic diameter, the tortuosity of the fluid flow paths, the maximum pore size and the fractal dimension of the cable cross-section. This theoretical correlation does not contain any fitting constants and each parameter has a clear physical meaning.

Thermal-hydraulic analysis of the DTT CS and PF pulsed coil performance during AC operation

The EU DEMO fusion reactor, now entering in its conceptual design phase, will have to tackle several challenges, such as the long plasma pulse duration and the exhaust of very high heat fluxes from the plasma. The Divertor Tokamak Test (DTT) facility, a compact superconducting tokamak under construction at ENEA Frascati, will address the power exhaust problem in DEMO by testing several DEMO-relevant divertor solutions and operation scenarios.The DTT superconducting coils are the subject of the analyses presented here: as the tokamak must be very flexible to face different plasma scenarios, the design of the magnet system must be proven to be robust. In particular, the thermal-hydraulic performance of the coils operated in pulsed mode, namely the Central Solenoid (CS) and the Poloidal Field (PF) coils, will be analyzed. The CS is composed of 6 modules, layer-wound with Nb3Sn Cable-in-Conduit Conductors (CICCs), while the PF coils are pancake-wound. The two PF coils closest to the machine axis are wound using Nb3Sn CICCs, while the others adopt NbTi CICCs. All the coils are cooled by supercritical He (SHe) in forced flow.The pulsed operation of these coils induces AC losses, eroding their temperature margin: a detailed thermal-hydraulic model of the DTT pulsed coils is developed here using the 4C code. The model is then applied to the simulation of the two reference plasma scenarios, a single null and a double null, computing the minimum temperature margin. Different cooling options (single- or double-pancake) are investigated for the PF coils, while the sensitivity to the coupling time constant value is assessed for the CS.

Impact of mechanical and thermal cycles at different operating conditions on the ITER toroidal field coil conductor performance

The impact of electromagnetic (EM) and thermal cyclic loading (referred as warm-up-cool-down—WUCD) on the Nb3Sn cable in conduit conductors for the toroidal field magnets of the ITER tokamak should be carefully assessed for the proper operation of the machine. The experience gained during the production phases of the TF conductors revealed that a degradation occurs when thermal and EM cyclic loading is applied at the nominal operating conditions of 10.78 T and 68 kA. Since at the beginning of the tokamak operation the nominal conditions will not be achieved, it is worth investigating to what extent lower EM loads affect the conductor performance, and the threshold of EM load that triggers the WUCD degradation. The aim of this investigation is to identify critical levels which determine the onset of this degradation, both in terms of thermal and EM cycles, and to study the performance of strands from different suppliers. For this assessment, a set of conductor samples were manufactured including all types of strands to be used in the TF coils of the ITER machine. These samples were subjected to a set of EM and thermal cycles representing in a realistic way the actual operation of the conductors in the ITER machine. The tests were performed in the SULTAN facility of the Swiss Plasma Center at working conditions ranging from 5.4 T–34 kA to 10.78 T–68 kA. This work presents the results obtained in these tests in terms of current sharing temperature (T cs), effective strain, rate of degradation at different working conditions and their implications for the prospected conductor performance in the machine.

Examination of design parameters affecting Nb3Sn CICC current sharing temperature using necessary condition analysis

The performance of Nb3Sn cable-in-conduit conductor (CICC) especially its current sharing temperature (T cs) is known to depend on the various parameters of the CICC such as the Nb3Sn strand characteristics, its cabling, the final geometry of the CICC cross section, the void fraction, the current distribution and other parameters. However, the performance has not been very predictable. The change in CICC T cs performance with repeated operation or electromagnetic loading is of particular concern. Furthermore, evaluation of T cs performance with respect to parametric variation of some CICC parameters has resulted in conflicting conclusions for different CICC designs. Thus, a more comprehensive review of published CICC T cs performance with respect to the corresponding CICC cable characteristics has been performed and is reported here for CICC samples whose designs and test results have been published during the past decade and more. Necessary condition analysis as well as linear regression is applied in the interpretation of the results which has found that the first stage cable twist pitch which has been the focus of much research is not that significant in determining the T cs performance but does affect T cs degradation. The direct effect of strand critical current on T cs performance was also found to be weak. On the other hand, void fraction was found to be significant in both determining T cs performance and limiting T cs degradation. Other correlations between parameters have also been studied. Overall, it seems that there has been appropriate targeting as well as overemphasis on some CICC parameters for improving Nb3Sn CICC T cs performance. No single parameter studied has been found to be a determining factor for high stable CICC T cs performance. Research on additional parameters is warranted.

Study of the ITER TF CICC Mechanical Behavior Under Cool-Down and Repetitive EM Loadings

High-current and high-field superconducting magnets sustain huge electromagnetic loads. On top of that, the high field constraint pushes towards the use of high-performing superconducting Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn strands, which are strain sensitive. In recent fusion magnets, the loads reach few hundreds of kN/m in static and sometimes cyclic regimes. These loads modify the strands local strain state and the overall mechanical properties of the cable. In this article, the finite element mechanical code MULTIFIL is applied to model an ITER Cable-In-Conduit Conductor (CICC) petal for the Toroidal Field (TF) coil under operating loadings. The local strain maps and the macroscopic mechanical behavior of the cable are analyzed for both thermal and electromagnetic loadings. A possible explanation for the initial drop of the current sharing temperature observed in some of the TF CICCs tested in the SULTAN facility (PSI, Villigen) is proposed. The impact of important model parameters, such as the void fraction and the friction coefficient between strands, on the mechanical behavior of the conductor is investigated.

Pressure Drop Measurements Performed on a Cable in Conduit Conductor (CICC) at a New Facility Developed to Study Thermo-Hydraulic Characterization

The recent trends in High Temperature Superconductor (HTS) development and HTS based applications have led to the exploration of various methods for achieving cryogenic temperature suitable for the superconducting cables and magnet characterization. An existing test facility at ITER-India has been augmented to deliver 80 K helium for testing, particularly for thermo-hydraulic characterization of superconducting cables. Such characterization is very important for the prediction of performance during operation to ensure adequate mass flow rate through the cable for heat load removal and to maintain the cable stability. The facility is capable of supporting the test on a superconducting short sample at 80 K with a mass flow rate up to 35 g/s and pressure up to 7 bar(a) using helium as a fluid. The functional acceptance of the developed test facility has been performed with the first experiment conducted on a short coiled shaped mock sample cable-in-conduit conductor (CICC). The longitudinal friction factor obtained from the measurement has been compared with the existing correlations for Reynolds number ranging between 50-400. The wound CICC short sample has been analyzed using Computational Fluid Dynamics (CFD) approach, considering the bundled region as a porous media to cover a wide range of various thermo-hydraulic and geometrical parameters involved in the test sample. A comparison of the experimental and the CFD simulation results is presented.

Suppression of the critical current degradation under the compressive stress on the internal reinforcement bronze processed Nb3Sn wire using Cu-Sn-In ternary bronze alloy matrix

Typical large current capacity conductors for fusion magnet are often manufactured by the Cable-In-Conduit (CIC) method. The CIC conductor is fabricated by twisting multiple Nb3Sn strands and pure Cu wires, and huge local stress/strain concentration alike the point contact was observed in each twisted Nb3Sn strand applied to transverse compressive load based on the electromagnetic force. These concentrations caused to the Nb3Sn filament breakage and it was also a major factor of the current sharing temperature (Tcs) and critical current density (Jc) degradation on CIC conductor. Mechanical strength improvement of the Nb3Sn strand is required to realize superconducting magnet under higher electromagnetic force for a future DEMO. Recently, we succeeded to fabricate the internal matrix reinforcement bronze processed Nb3Sn multifilamentary wire using a Cu-Sn system ternary alloy matrix containing Indium (In) element.In this study, critical current (Ic) property with the in-situ compressive stress on the internal matrix reinforcement Nb3Sn wire was evaluated. The compressive stress obtained to the Ic deterioration of 5% on the internal matrix reinforcement Nb3Sn wire was confirmed at 150 MPa, and this stress was 3 times higher than Nb3Sn wire without reinforcement. We found that the internal matrix reinforcement was one of the great advantage methods to suppress the Ic degradation by the compressive stress.

Performance test and analysis of the first large-scale cable-in-conduit conductor with high J c Nb3Sn strand for fusion reactor

The Comprehensive Research Facility for Fusion Technology (CRAFT) project has been launched in 2019, for developing the essential engineering technologies for Chinese Fusion Engineering Testing Reactor (CFETR). Within this project, a full-size toroidal field (TF) coil will be built as the prototype coil for CFETR. Based on design of CFETR magnet system, the TF coil will operate at 95.6 kA in a peak field of 14.5 T. The high-J c Nb3Sn strand is taken into consideration due to the critical current density of ITER-grade Nb3Sn is too low at 14.5 T. Considering that it will be the first time to apply the high-J c Nb3Sn strand in the large-scale cable-in-conduit conductor (CICC) for fusion magnet, a conductor sample made of high-J c Nb3Sn strand with short twist pitch (STP) cable pattern was manufactured in ASIPP and tested in SULTAN facility, to investigate the feasibility. The test campaign focuses on the impact of cyclic electromagnetic (EM) loading and warm-up cool-down (WUCD) to the performance of the conductor, the strain distribution of the conductor before and after EM cycles was measured by inductive method to make a deeper insight of the conductor performance evolution. AC losses tests have also been carried out, providing relevant information for further coil design.

Extensive Analyses of Superconducting Cables 3D Geometry With Advanced Tomographic Examinations

In the framework of activities embedding magnet design and associated R&D activities and relying on Cable in Conduit Conductors (CICC) technology, the singularity of the concept can rise some challenges versus their modelling in operation. Indeed, CICC includes thousands of superconducting strands, twisted together under a multi-staged scheme that also includes deformation by cabling and compaction during manufacture. This causes the accurate predictability of strands position in CICC extremely difficult, while it can be a key element for modelling their performances in operation. As a matter of fact the coupling losses rely on the CICC capability to establish inter-strand shielding currents, driven by the inter-strand contacts mapping which are only accessible in prediction via accurate 3D strand trajectories geometry. Same applies for prediction of CICC mechanical properties (deformation of Nb3Sn strands) and hydraulic properties (helium coolant force-flowed between the strands), making those investigations of high added-value. In this context, INFLPR installed a new set-up dedicated to micro-tomography to examine CICCs with high resolution, and get an overall overall 3D overview regarding picture of the strands location. CEA and INFLPR further developed a post-processing method to reconstruct the strands trajectories with high accuracy. The measurement and data analysis workflow was applied to two CEA middle-size CICCs with variable void fraction from which contact statistics were issued. The obtained database was exploited to reconstruct equivalent 3D resistive network. Above applications using CICC topology database are discussed, compared to experimental AC losses database issued by CEA and interpreted considering the CEA model COLISEUM. The outcomes are discussed and subsequent guidelines for future work presented.

Development of the H4C Model of Quench Propagation in the ENEA HTS Cable-In-Conduit Conductor

Experiments on quench propagation in high-current, high field High Temperature Superconducting (HTS) Cable-In-Conduit Conductors (CICCs) for fusion applications are forthcoming. Among the conductor designs to be tested, there is the ENEA slotted core proposal. In order to support the design of the samples and plan the diagnostics to be employed, a 1D multi-region thermal-hydraulic and electric model of the samples has been developed using the H4C code. After an experimental electric characterization, the model is applied to the simulation of quench propagation in the samples. The simulations show how current redistributes among the tapes and the slots. Additionally, they show that the quench protection strategy is suitable to prevent too high hot-spot temperatures.