Central Solenoid Model Coil Research Articles

Status of the ITER magnets

The overall programme objective of ITER is to demonstrate the scientific and technologicalfeasibility of fusion power production for peaceful purposes. ITER would accomplish thisobjective by demonstrating controlled burn of deuterium tritium plasmas, andby demonstrating essential technologies (of a reactor) in an integrated system.The magnetic coils of ITER are all made of superconducting conductors. A comprehensiveresearch and development programme has been carried out to demonstrate the feasibility ofthe magnetic system. The major elements of this programme have been the construction ofthe toroidal field and central solenoid model coils. The testing of these model coils hasproduced all the necessary information to finalize the design of the ITER magneticsystem. In addition, a poloidal field conductor insert is being manufactured and willbe tested in 2006 at Naka (J) to confirm the design of the poloidal field coils.

Measurements of the microstructural, microchemical and transitiontemperature gradients of A15 layers in a high-performanceNb3Sn powder-in-tube superconducting strand

In the past 5 years there has been a remarkable increase in layer critical current density,Jc, forNb3Sn.At ∼5000 A mm−2 (12 T, 4.2 K), the average layer critical current density is now double that of the bestITER central solenoid model coil strand. The improvements in critical currentdensity are a result of increased Sn content in the superconducting A15 phase andbetter compositional homogeneity, and perhaps other reasons not yet understood.The first indication that such large increases in critical current density of theNb3Sn layer were possible was in the powder-in-tube strands produced by Shape Metal Innovation.The design of these strands allows us to accurately measure the compositiongradient as well as the gradient in critical temperature across the A15 layer, and wereport these measurements here for a wide range of heat treatments. InductiveTc measurements were used to measure the radial gradient ofTc, while specific heat measurements revealed the position-insensitive inhomogeneity of the A15layer. Composition gradients were measured by energy dispersive x-ray spectroscopy in a fieldemission scanning electron microscope. The three characterizations show that the composition andTc gradient is quite shallow near the central Sn source and only becomes steep adjacent to the Nbsheath of each filament, a result that is beneficial to maximizing the fraction of the layer with highTc. This strongnonlinearity of the Tc and composition gradient means that excellent properties are obtained, in spite of thebuilt in composition gradients inevitable in any filamentary design of composite.The gradients are much smaller than reported for bronze-route conductors buthigher than the overall layer gradient found in the latest generation of high criticalcurrent density internal Sn design strands. Coupled to recent modelling simulations,our data show the great value of high-Sn designs in developing high layerJc valuesin Nb3Sn conductors.

低温工学に関わる真空の物理と技術 核融合炉用超伝導コイル試験装置の真空漏れ探索技術の開発

Japan Atomic Energy Agency (JAEA) has developed leak detection technology for liquid helium temperature experiments in large-scale superconducting magnet test facilities. In JAEA, a cryosorption pump that uses an absorbent cooled by liquid nitrogen with a conventional helium leak detector, is used to detect helium gas that is leaking from pressurized welded joints of pipes and valves in a vacuum chamber. The cryosorption pump plays the role of decreasing aerial components, such as water, nitrogen and oxygen, to increase the sensitivity of helium leak detection. The established detection sensitivity for helium leak testing is 10-10 to 10-9 Pam3/s. A total of 850 welded and mechanical joints inside the cryogenic test facility for the ITER Central Solenoid Model Coil (CSMC) experiments have been tested. In the test facility, 73 units of glass fiber-reinforced plastic (GFRP) insulation break are used. The amount of helium permeation through the GFRP was recorded during helium leak testing. To distinguish helium leaks from insulation-break permeation, the helium permeation characteristic of the GFRP part was measured as a function of the time of helium charging. Helium permeation was absorbed at 6 h after helium charging, and the detected permeation is around 10-7 Pam3/s. Using the helium leak test method developed, CSMC experiments have been successfully completed.

Results of magnet tests as a basis for the design of ITER magnets

To ensure the availability of the ITER magnets an extensive development programme has been executed to demonstrate that the technology is available on industrial scale. This programme included two major projects, firstly the development and testing of a central solenoid model coil (CSMC) with additional testing of high field inserts and secondly the development and testing of a toroidal field model coil (TFMC), both supported by investigations on conductors, structural material and critical components. The model coils have been successfully developed, fabricated and tested in dedicated test facilities at JAERI/Naka (CSMC) and Forschungszentrum Karlsruhe (TFMC), respectively. The tests confirmed that the design criteria are met by the model coils and the ITER coils can be built on this basis. Of course, the tests showed also in which areas improvements should be implemented. One concerns the Nb 3Sn conductor. To guarantee the defined margin of the current sharing temperature, advanced conductor strands with higher critical current density than those used in the model coils have to be used and are available now. The codes for mechanical behaviour, ac loss predictions and thermohydraulic behaviour gave rather good agreement with the measurements, but can still be improved. Quality assurance issues concern high voltage properties and leak tightness.

Axial tensile stress–strain characterization of ITER model coil typeNb3Sn strands in TARSIS

For a few years there has been an increasing effort to study the impact of (bending) strainon the transport properties of superconducting wires. As the stress distribution,originated by differences in the thermal expansion and electromagnetic load, isthe driving factor for the final strains, the axial and transverse stiffness of thestrand play a crucial role in the final performance. Since the strain state of theNb3Sn filaments in strands determines the transport properties, basic experimentalstress–strain data are required at the strand level for accurate modellingand analysis and eventually for optimizing cable and magnet design. Weperformed axial tensile stress–strain measurements on several types ofNb3Sn strands used for the manufacture of the International Experimental Thermonuclear Reactor(ITER) central solenoid and toroidal field model coils and a powder-in-tube processed wire.In total 48 wire samples were tested at boiling helium, boiling nitrogen and at roomtemperature. We present the computation of the stress–strain characteristic with astraightforward 1D model using an independent materials database, obtaining a goodagreement with the experimental results.The details from the take-off origin of the measured stress–strain curves are discussed andthe data are evaluated with respect to some commonly used functions for fittingstress–strain curves. The measurements are performed in the new setup TARSIS (testarrangement for strain influence on strands). A double extensometer connected to thesample enables us to determine the strain level whereas a load cell is used to monitor thestress level.For higher levels of applied stress (100 MPa), we found typically a higher strainfor bronze route wires compared to a powder-in-tube and internal tin typeof strand. Stress–strain results are essential to assess more accurately theimpact of thermal and electromagnetic induced stress on the strain state of theNb3Sn filaments for wires from various manufacturing processes.

The Test Facility for the EAST Superconducting Magnets

A large facility for testing superconducting magnets has been in operation at the Institute of Plasma Physics of the Chinese Academy of Sciences since the completion of its construction that began in 1999. A helium refrigerator is used to cool the magnets and liquefy helium which can provide 3.8 K-4.5 K, 1.8 bar-5 bar, 20 g/s-40 g/s supercritical helium for the coils or a 150 L/h liquefying helium capacity. Other major parts include a large vacuum vessel (3.5 m in diameter and 6.1 m in height) with a liquid nitrogen temperature shield, two pairs of current lead, three sets of 14.5 kA-50 kA power supply with a fast dump quench protection circuitry, a data acquisition and control system, a vacuum pumping system, and a gas tightness inspecting devise. The primary goal of the test facility is to test the EAST TF and PF magnets in relation to their electromagnetic, stability, thermal, hydraulic, and mechanical performance. The construction of this facility was completed in 2002, followed by a series of systematic coil testing. By now ten TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a model coil of the PF large coil have been successfully tested in the facility.

Thermo-Hydraulic Analysis of the KSTAR Central Solenoid Model Coil

The Korea Superconducting Tokamak Advanced Research (KSTAR) Central Solenoid Model Coil (CSMC) has been developed to validate the design of KSATR CS coil. The thermo-hydraulic characteristics were analyzed for the KSTAR CSMC. The major thermo-hydraulic parameters of the coil are AC losses, strand temperature, coolant temperature, pressure drop and temperature margin. A numerical code has been developed for the thermo-hydaulic analysis of the KSTAR CSMC according to the operating conditions. In this paper, the description of the thermo-hydraulic analysis models and analysis results of CSMC are presented. The total energy deposition due to AC losses in CSMC winding will induce a flow reversal of supercritical helium. The minimum temperature margin is estimated approximately 0.5 K. The maximum temperature in CSMC winding is about 7.6 K.

Quench Detection Using Pick-Up Coils for the ITER Central Solenoid

At least two independent methods for quench detection should be used for the magnets of the ITER size, because of the potential severe impact of a failure. The voltage tape method and the flow meter method, which were used in the CS model coil (CSMC), are both considered for the ITER Central Solenoid (CS). The voltage tap method is primary due to its quick response. The ITER CS consists of six pancake wound modules, which are operated with individual operating current patterns in ac mode. The induced voltage in the windings must be compensated to detect the voltage due to any normal transition during the pulse operation. We have investigated the optimum configuration for pick-up coils (PC) for compensation. The results of simulations show that the compensated voltages are very low (70 mV) compared with the inductive voltage and adequate normal voltage sensitivity is obtained. The hot spot temperature in the CS during the operation was estimated from the simulation and the experimental measurements of the CSMC quench. The hot spot temperature estimated is 128-144 K, lower than the ITER design criterion (150 K). It is shown that the detection system using the PCs could be designed with a high enough detection sensitivity.

Design and Manufacture of a Full Size Joint Sample (FSJS) for the Qualification of the Poloidal Field (PF) Insert Coil

Within the framework of the European Fusion Programme a FSJS has been designed and manufactured by European Industry using PF coil NbTi superconductor manufactured and supplied by the Russian Federation as part of the R&D for the PF Conductor Insert (PFCI) coil. In addition to the superconductor, this sample contains a number of unique features. In contrast to other samples previously manufactured in Europe, the FSJS has used the Central Solenoid Model Coil (CSMC) joint design with NbTi conductor and a thick square jacket. One leg of the FSJS has had the conductor and sub-petal steel wraps removed before jacketing to assess the difference in the conductor performances between the two configurations. This paper will report on the development and manufacture of the FSJS, in particular the use of the Central Solenoid jacketing and swaging tools for compaction of the conductor and swaging of the joints, the preparation and qualification of the manufacturing route for the joint, the jacketing of a special length without conductor wraps and the high level of instrumentation required for the testing of the joint. The sample has been instrumented with more sensors than any other previous European sample, including temperature sensors, a large number of voltage taps for quench detection and Tcs measurements, quadrupoles to detect uneven voltage distribution, hall arrays for current distribution measurements and saddle coils.

Effects of Bending on Cracking and Critical Current of<tex>$rm Nb_3rm Sn$</tex>ITER Wires

Critical current (I/sub c/) measurements and microscopic crack investigations were performed on short samples of ITER (International Thermonuclear Experimental Reactor) wires, designed for the CSMC (Central Solenoid Model Coil), that were bent to curvatures ranging up to 0.3% at room temperature. I/sub c/ was found to improve significantly with bending. Microstructural investigation revealed that IGC and VAC samples were nearly crack-free at all bend strains while Furukawa samples had significant cracking due to polishing and unrelated to bending.

Transverse Stress Effects in ITER Conductors

The International Thermonuclear Experimental Reactor (ITER) Central Solenoid and Toroidal Field Model Coils (CSMC, TFMC) and the CSMC Insert Coils using Nb/sub 33/Sn had unexpectedly low current-sharing temperatures and cable "n-values." This paper argues that the anomalous effects can be explained by the high longitudinal strain due to transverse Lorentz loads. Simulated results of earlier experiments with A15 monoliths indicate that the transverse stress "multiplier" is actually a longitudinal strain multiplier due to transverse loading. Irreversible cyclic degradation effects in the cables with the lowest thermal contraction mismatches are explained by showing that a modest precompression is required to prevent tensile cracking of a significant number of individual filaments. Examination of filament breakage due to bending shows the possibility of a well-defined threshold for irreversible damage.

Preparation of the ITER Poloidal Field Conductor Insert (PFCI) Test

The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in the EU and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, /spl sim/45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of /spl sim/6 T, and it will be cooled by supercritical He at /spl sim/4.5 K and /spl sim/0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure gauges, strain and displacement sensors. The test program will be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermal-hydraulic characterization. Here we give an overview of the preparatory work toward the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far.

CICC Joint Development and Test for the Test Facility

The superconducting joint of the NbTi Cable-in -conduit Conductor (CICC) has been developed and tested on the magnet test facility at Institute of Plasma Physics, Chinese Academy of Sciences. The CICC is composed of (2NbTi+1Cu)×3×3×(6+1tube) strands each with 0.85 mm in diameter, which has been developed for a central solenoid model coil. The effective length of the joint is about 500 mm. There have been two common fabrication modes, one of them is to integrate the 2 CICC terminals with the copper substrate via lead-soldering, and the other is to mechanically compress the above two parts into an integrated unit. In the current range from 2 kA to 10 kA the joint resistance changes slightly. Up to now, 11 TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a large PF model coil of PF large coil have been completed via the latter joint in the test facility.

Progress in the Superconducting Magnet Technology Through the ITER CS Model Coil Programme

The world's largest high-field pulsed super conducting magnet, the CS model coil, was manufactured and tested by the international and participant teams for the verification of new design principles and fabrication concepts proposed for the ITER CS. State-of-the-art technologies were developed through the CS model coil program including high-performance Nb/sub 3/Sn strand, 46 kA cable-in-conduit conductor, and coil fabrication using "wind, react, insulate and transfer" method that enables Nb/sub 3/Sn to be applicable to a large magnet with mechanically rigid and high-voltage insulated windings. The CS model coil reached its design point of 13 T and 46 kA and demonstrated the pulsed operation of 0.6 T/s to 13 T, achieving all the development goals. The experiments of the CS model coil and three single-layer inserts have provided detailed knowledge on the performances of the coils and conductors including the hydraulic behavior, AC losses and critical currents. In particular, significant effects of large electromagnetic forces acting on the conductor were observed in these performances and the movement of cable inside the conductor was identified to be important. Results and experiences from the program are not only useful for the design optimization of the ITER magnets, but also applicable to other areas such as high energy physics.

Performance of an ITER CS1 Model Coil Conductor Under Transverse Cyclic Loading up to 40,000 Cycles

The large currents in the cable-in-conduit conductors (CICC) destined for the high field magnets in the International Thermonuclear Experimental Reactor (ITER), cause huge transverse forces on the strands compressing the cable against one side of the conduit. This load causes transverse compressive strain in the strands at the crossovers contacts. Moreover, the strands are also subjected to bending and contact surfaces micro-sliding, which results into friction and anomalous contact resistance versus force behavior. Three Nb/sub 3/Sn central solenoid model coil (CSMC) conductors were tested previously in the Twente Cryogenic Cable Press up to 40 cycles with a transverse peak load of 650 kN/m. This press can transmit a variable (cyclic) transverse force directly to a cable section of 400 mm length at a temperature of 4.2 K (or higher). To explore life-time cycling, we tested a CSMC Nb/sub 3/Sn conductor up to 40,000 cycles. The coupling loss and the associated interstrand resistance between various strands and strand bundles are measured at various loads. The force on the cable and the displacement are monitored in order to determine the effective cable Young's modulus and the mechanical heat generation. Some aspects of strand deformation in CICC's are discussed. The test results are discussed in view of previous press results and data extracted from the ITER model coil tests.

Proposals for the Final Design of the ITER Central Solenoid

The ITER central solenoid (CS) is 12 m high and 4 m in diameter. The CS consists of a stack of 6 electrically independent modules to allow control of plasma shape. The modules are compressed vertically by a pre-compression structure to maintain contact between modules. The CS conductor is CIC conductor with Nb/sub 3/Sn strands and a steel conduit. The CS model coil and insert coil test results have shown that the conductor design must be modified to achieve an operation margin. This required either to increase the cable diameter or to use strand with a higher current capability. A bronze-process (NbTi)/sub 3/Sn strand is proposed to achieve a higher critical magnetic field. A square conduit with a high Mn stainless steel is proposed as it can satisfy fatigue requirements. The inlets are in the high stress region and any stress intensification there must be minimized. The preload structure is composed of a set of tie-plates, flanges, buffer plates, and wedges. These design proposals satisfy all ITER operational requirements.

Analysis and Interpretation of<tex>$rm T_rm cs$</tex>Tests of the ITER Toroidal Field Conductor Insert

The toroidal field conductor insert (TFCI) coil was tested at 46 kA in the 13 T background magnetic field generated by the central solenoid model coil. The single layer TFCI coil uses a Nb/sub 3/Sn cable-in-conduit conductor in a titanium jacket. The object of this work is to verify the data self-consistency, and to apply a thermal, hydraulic and electrical simulation to derive the strand performance in the cable. We present the results of the analysis in support of the interpretation of a subset of current sharing temperature measurements. After estimation of the n-value in the voltage-current equation, we perform the simulations using the total longitudinal strain as the only fitting parameter; the resulting value is much larger than the value expected from simple thermal contraction considerations.

Performance of Joints in the CS Model Coil and Application to the Full Size ITER Coils

Cable-in-conduit conductors (CICC) consisting of about 1000 Nb/sub 3/Sn strands with an outer diameter of about 0.8 mm, have been designed for the TF and CS coils of the ITER. The rated currents of these coils are 68 and 40 kA, respectively. Two joint types (Butt and Lap) were developed during the CS Model Coil project. The performance of these joints was evaluated during the operating tests and satisfactory results were obtained. The joints of the TF coils are located outside of the winding in a region where the magnetic field is about 2.1 T, a very low value as compared to the maximum field of 11.8 T at the winding. The CS joints are located at the coil outer diameter and embedded within the winding pack to minimize the space occupied around the coil. The maximum fields at the CS joint and winding are 3.5 and 13 T, respectively. For the TF coils and the CS, the joints are cooled in series with the conductor at the outlet. The maximum allowable temperature increase due to the joule heating in the joints is 0.15K to reduce the heat load on the refrigerator. The stability of these joints at the ITER operating condition was analyzed by using the experimental data. The analysis shows that both joint types are applicable to the full size ITER coils.

Measurement of High Frequency Impedance of ITER CS Model Coil and the Simulation Analysis of Internal Resonance Phenomenon

This paper describes the internal resonance phenomenon of a large scale superconducting coil. The frequency dependence of electric impedance defined at the coil terminal and the voltage distribution of the layer winding were measured at the ITER CS model coil under the superconducting state in advance of the pulse operation test. As a result, large change of the impedance and unbalance of the layer voltage were observed. The AC circuit analysis was carried out in order to evaluate the measurement results and to calculate the voltage of layer insulator. It was concluded that the higher voltage, which may appear on the layer insulation, is about a half of the coil terminal voltage in high frequency.

Ｎｂ３Ｓｎ線材の現状と将来展望―発見から５０年を記念して―核融合用Ｎｂ３Ｓｎ導体の開発

The progress of Nb3Sn conductor development for fusion power applications at JAERI is described. The superconductor development was carried out in connection with the research and development for four fusion coil development projects: Phoenix, Cluster test, Demo Poloidal Coil and ITER-EDA. In the Phoenix project, the reliable production of multi-filamentary Nb3Sn strands was established during the development work towards realizing a 18 T coil, and an 18 T generation coil (BC-18) with multi-filamentary Nb3Sn strands was successfully operated at 4.2 K for the first time in the world. In the Cluster Test project, a pool-cooled Nb3Sn conductor with a high heat-flux surface, and the first 1-m class bore Nb3Sn coil (TMC) in the world was constructed and operated up to 12 T. In the Demo Poloidal Coil project, which was constructed to demonstrate the possibility of a poloidal coil, a forced-cooled Nb3Sn conductor was developed. The Nb3Sn coil (DPC-EX) constructed with it was pulse-operated up to 7 T in 0.5 s. Finally, in the ITER-EDA project a CICC-type 13 T-46 kA Nb3Sn conductor was developed. A Central Solenoid Model Coil (CSMC) was fabricated using it and successfully operated up to 13 T with 0.4 T/s.