International Thermonuclear Experimental Reactor Conductor Research Articles

Statistics of Test Results for the ITER Nb3Sn and NbTi Conductors in the SULTAN Facility

The series production of Nb3Sn and NbTi cable-inconduit conductors (CICC) for the winding of various coils in the International Thermonuclear Experimental Reactor (ITER) magnet system is being completed in 2015. The tests of the ITER conductor samples are running in the SULTAN test facility in Villigen, Switzerland, in the scope of a framework contract, flanked by six bilateral agreements between IO and the Domestic Agencies (DAs) of China, EU, Korea, Japan, Russian Federation, and USA. The key acceptance parameter of the tests for the conductor samples made from the series production ITER CICCs is the current sharing temperature T <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">cs</sub> at the nominal operating field and current after applied electromagnetic cyclic loadings at the nominal field and sinusoidal swing of conductor current from zero to nominal one. The T <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">cs</sub> is defined as the maximum temperature at which the conductors operate (nominal current and field) before developing an electric field of 10 μV/m. This paper gathers the latest test results of the ITER series production CICC. The qualified conductor units will be used in the winding of the toroidal field and poloidal field coils, as well as the central solenoid. The test results for the series production conductors are compared with the results of the conductors tested at the early stages of their development. The performance variation among the suppliers and the performance evolution since the start of the conductor production are discussed as well.

NbTi Strands Verification for ITER PF CICC Process Qualification of CNDA

China is in charge of most of Poloidal Field (PF) conductors production for the International Thermonuclear Experimental Reactor (ITER). The execution for PF conductors shall be in three main phases. According to ITER Procurement Arrangement (PA), the Domestic Agency (DA) shall be required to verify the room and low temperature acceptance tests carried out by the strand suppliers. As the reference laboratory of Chinese DA (CNDA), the superconducting strands test laboratory of Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) was undertaking the task of strands verification for ITER conductors. The verification test includes: diameter, Nickel plating thickness, copper-to-non-copper volume ratio, twist pitch direction and length, standard critical current (IC) and resistive transition index (n), residual resistance ration (RRR), and hysteresis loss. 48 NbTi strands with 7 billets were supplied for the PF Cable-In-Conduit Conductor (CICC) process qualification. In total, 54 samples were measured. The verification level for PF CICC process qualification was 100%. The test method, facility and results of each item are described in detail in this publication.

Electromagnetic and Stress Analysis of 50 kA Superconducting Transformer for ITER Conductor Test Facility

A superconducting transformer was designed for the correction coils (CC) conductor test facility in the Institute of Plasma Physics, the Chinese Academy of Sciences (ASIPP), for validating the CC conductor of the international thermonuclear experimental reactor (ITER). The maximum current of the superconducting transformer is 50 kA. The conductor samples could be tested at different magnetic fields and cooling conditions similar to the operational condition of ITER. During normal operation, the primary and secondary windings of the superconducting transformer will withstand the high electromagnetic (EM) stress caused by the high current. Integrated analysis of EM and structural fields can ensure these components not to encounter any over-stress problem under normal operational condition. A coupled model is proposed to combine the EM and structural analyses. A detailed finite element (FE) model and EM-structural coupled model are presented and the numerical results show that the stress of the transformer windings is within the allowable limits.

Investigation on Thermal Strain in Full Size ITER Conductors

The TF magnetic system of the international thermonuclear experimental reactor (ITER) will be composed of 18 superconducting coils assembled in a torus. The design of these large coils relies on the use of a circular cable-in-conduit conductor (CICC) composed of a cable of about one thousand twisted Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and copper strands embedded in a thin steel jacket. The different thermal contractions of the materials composing this conductor during the cool down from the Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn heat treatment reaction temperature (about 920 K) to the operating temperature (about 5 K) result in a compressive strain on the superconducting Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn filaments which leads to a degradation of the conductor critical current. The qualification of the ITER conductor prototypes will be performed by testing short straight full size samples in the unique European test facility SULTAN at PSI (Switzerland). A comparison of the performance of the toroidal field model coil (TFMC) conductor in a short straight sample and in the coil has shown indications that the sample performance is better, apparently because the thermal strain in Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn is lower. It looks possible that this could have been caused by a slippage between the cable and the jacket in the joints at the sample ends during the heat treatment. In this case, these kind of short straight samples could not be relevant for conductor qualification. An investigation on the thermal strain by destructive examination of different TFMC conductor samples, including one leg of the TFMC short sample tested in 1999 in the SULTAN facility, was performed at CEA Cadarache. The paper reports on the different strain measurements, proposes an analytical approach of the results in order to conclude on the relevance of straight samples.

Transverse load optimization inNb3Sn CICC design; influence of cabling, void fraction and strand stiffness

We have developed a model that describes the transverse load degradation inNb3Sn CICCs, based on strand and cable properties, and that is capable of predicting how suchdegradation can be prevented.The Nb3Sn cable in conduit conductors (CICCs) for the International Thermonuclear ExperimentalReactor (ITER) show a significant degradation in their performance with increasingelectromagnetic load. Not only do the differences in the thermal contraction of thecomposite materials affect the critical current and temperature margin, but mostlyelectromagnetic forces cause significant transverse strand contact and bending strain in theNb3Sn layers.Here, we present the model for transverse electro-magnetic load optimization (TEMLOP)and report the first results of computations for the ITER type of conductors, based on themeasured properties of the internal tin strand used for the toroidal field model coil(TFMC). As input, the model uses data describing the behaviour of single strandsunder periodic bending and contact loads, measured with the TARSIS set-up,enabling a discrimination in performance reduction per specific load and strand type.The most important conclusion of the model computations is that the problem of thesevere degradation of large CICCs can be drastically and straightforwardly improvedby increasing the pitch length of subsequent cabling stages. It is the first timethat an increase of the pitches has been proposed and no experimental data areavailable yet to confirm this beneficial outcome of the TEMLOP model. Largerpitch lengths will result in a more homogeneous distribution of the stresses andstrains in the cable by significantly moderating the local peak stresses associatedwith the intermediate-length twist pitches. The twist pitch scheme of the presentconductor layout turns out to be unfortunately close to a worst-case scenario.The model also makes clear that strand bending is the dominant mechanism causingdegradation. The transverse load on strand crossings and line contacts, abbreviated ascontact load, can locally reach 90 MPa but this occurs in the low field area of the conductorand does not play a significant role in the observed critical current degradation. The modelgives an accurate description for the mechanical response of the strands to a transverseload, from layer to layer in the cable, in agreement with mechanical experiments performedon cables.It is possible to improve the ITER conductor design or the operation margin, mainly by achange in the cabling scheme. We also find that a lower cable void fraction andlarger strand stiffness add to a further improvement of the conductor performance.

Influence of the magnetic field profile on ITER conductor testing

We performed simulations with the numerical CUDI-CICC code on a typical short ITER(International Thermonuclear Experimental Reactor) conductor test sample of dual legconfiguration, as usually tested in the SULTAN test facility, and made a comparison withthe new EFDA-Dipole test facility offering a larger applied DC field region. The newEFDA-Dipole test facility, designed for short sample testing of conductors for ITER, has ahomogeneous high field region of 1.2 m, while in the SULTAN facility this region is threetimes shorter. The inevitable non-uniformity of the current distribution in the cable,introduced by the joints at both ends, has a degrading effect on voltage–current(VI) andvoltage–temperature (VT) characteristics, particularly for these short samples. This can easily result in anunderestimation or overestimation of the actual conductor performance. A longer appliedDC high field region along a conductor suppresses the current non-uniformity by increasingthe overall longitudinal cable electric field when reaching the current sharing mode. Thenumerical interpretation study presented here gives a quantitative analysis for a relevantpractical case of a test of a short sample poloidal field coil insert (PFCI) conductor inSULTAN. The simulation includes the results of current distribution analysis fromself-field measurements with Hall sensor arrays, current sharing measurements andinter-petal resistance measurements. The outcome of the simulations confirms that thecurrent uniformity improves with a longer high field region but the ‘measured’VI transition is barely affected, though the local peak voltages become somewhat suppressed. It appearsthat the location of the high field region and voltage taps has practically no influence on theVI curve as long as the transverse voltage components are adequately cancelled. In particular,for a thin conduit wall, the voltage taps should be connected to the conduit in the form ofan (open) azimuthally soldered wire, averaging the transverse conduit surface potentialsinitiated in the joints.

The Effect of Inter-bundle Resistive Barriers on Coupling Loss, Current Distribution and DC Performance in ITER Conductors

The role of inter-bundle resistive barriers (metal sheet wraps), introduced to reduce the inter-bundle coupling loss in multistage cabled Cable-In-Conduit Conductors (CICC) for the International Thermonuclear Experimental Reactor (ITER) is evaluated, based on results gained recently on short sample experiments in the Twente Cable Press and SULTAN. The obvious benefit of limiting the inter bundle coupling loss unavoidably goes together with impeding the redistribution of nonuniform currents in the coil winding introduced at the terminations, as well as reduction of the heat exchange between the bundles. Six-element numerical electromagnetic code simulations are presented that qualitatively explain the effect of wraps on the DC performance, strongly depending on the testing geometry. The computations illustrate that wraps can reduce the DC performance in short sample tests. At the same time simulations of the Poloidal Field Coil Insert (PFCI), with a winding length of 50 m, have shown that omitting sub-stage wraps, can even degrade the DC performance of coils due to the short current transfer length in combination with current nonuniformity causing peak voltages in the most overloaded petals

Effect of self-field and current non-uniformity on the voltage–temperature characteristic of the ITER central solenoid insert coil by numerical calculations

For the optimisation of a magnet design with cable-in-conduit conductor (CICC) technology it is essential to comprehend the scaling of the critical current from the separate strand characteristics to the finally assembled cable performance in a coil. Several model coils have been tested in the framework of research for the International Thermonuclear Experimental Reactor (ITER). At present, the scaling of the critical current from the strand to the full cable performance and the apparent decrease of the n-index from strand to cable in the voltage–current curves is not understood. It is important to recognize the mechanisms behind this phenomenon in relation to the cost of the superconducting strand, which is significant in the manufacture of the magnets. Therefore, basic phenomena like the cable conductor self-field, the current unbalance introduced by the non-uniformity of the joints and a possible reversible or irreversible degradation of the voltage current characteristic of a strand during cable manufacture or electromagnetic loading of the magnet have to be considered. The voltage–current characteristic of the strand is extensively explored for the relevant range of magnetic field, temperature and axial strain space. Accordingly a numerical six-element network model is developed to simulate the conditions and behaviour of the last stage cable elements of a full-size ITER conductor. The experimental data, mainly in terms of voltage–current ( VI) or -temperature ( VT) characteristics, are obtained on the central solenoid insert coil (CSIC) experiment performed in Naka (Japan) in the framework of the research for ITER. The numerical model, which is briefly introduced, is used to study the cable performance by using experimentally obtained cable parameters like inter-strand (and bundle) contact resistance, strand critical current data as a function of magnetic field, temperature and applied axial strain, and external cable self-field measurements by Hall sensors for reconstruction of the current non-uniformity. The effect of a current redistribution due to the cable self-field on the voltage–temperature curve is calculated in correlation with the transverse resistance between the strands and last cabling stage bundles (petals). A realistic unbalanced current distribution is established by introducing non-uniform joints at the extremities of the CS-insert cable. It appears that the cable self-field effect hardly gives any change in the shape of the VT curve but merely a shift towards lower temperature giving a reduction of the current sharing temperature T cs (10 μV/m) of <0.1 K. For typical CICCs with Cr-coated Nb 3Sn strands, there is practically no current redistribution due to the cable self-field, because of the high inter-strand contact resistance. An unbalanced current distribution also gives an earlier voltage rise in the VT curve, mainly at low levels of the electric field. At a 10 μV/m criterion practically no reduction of the T cs (<0.1 K) is found by the numerical simulation. However, in the CSIC the experimentally obtained overall reduction in T cs from strand to cable is 0.7 K for an operating current of 40 kA at 12.5 T background field. According to the results of the numerical simulation, the cable self-field effect and the non-uniform current distribution, which is unavoidably caused by the joints, cannot explain the early voltage rise and low n-index in the VT curve of the CS-insert coil. It is very likely that electromagnetic forces play a role in causing reversible degradation in critical current or even irreversible due to strand (filament) damage. Neither can it be excluded that strand deformation during cabling has an impact on the final conductor performance as well. Therefore additional effort is required in detailed 3D modeling of the possible strand deformations inside a cable and the impact it has on the strand performance by experimental verification on strand level.

ケーブル・イン・コンジット型超伝導導体の圧力損失及び結合損失に対する電磁力の影響

In the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER), a Central Solenoid Model Coil (CSMC) and a CS Insert Coil (CSIC) have been tested successfully. The CSIC conductor consists of 1, 152 superconducting strands bundled on a central cooling channel. As interesting phenomena in the CSIC experiment, it was observed that a pressure drop of the CSIC decreased by about 12% during a current-carrying operation at 40kA, and coupling losses indicated an operating current dependence. It is considered as a hypothesis that an electromagnetic force causes a compressive deformation of superconducting cable in a jacket and that a new flow path was then generated between cable and jacket. Therefore it is also considered that the decreasing of contact resistance between strands as a result of the electromagnetic force derives an increase of coupling losses in the conductor. A pressure drop calculation model with a gap generated by electromagnetic force is constructed. The gap is estimated to be about 1.4mm at nominal operating conditions (13T, 44.3kA). From this calculation, a void fraction as a function of electromagnetic force is evaluated during the current-carrying operation of CSIC. The coupling time constant (nτc) as a function of void fraction is then calculated from the coupling loss measurement result during the pulsed operation of CSMC and CSIC. The evaluated nτc is about 24ms and is close to nτc of 20-30ms of a heat treated short sample having a history of exposure to the electromagnetic force. We used the evaluated nτc as a function of electromagnetic force to calculate the coupling losses, which varied from 24ms to about 50ms during pulsed current operation. These results show a good agreement with measured coupling losses, depending on coil current. To reduce the possibility of strand damage as a result of cable movement, we also here proposed that the void fraction of real ITER conductor should be smaller than that of CSIC, and it is preferable that the void fraction is about 34.5%. In this paper, the quantitative explanation of coupling loss change under the electromagnetic force is described from the viewpoint of the pressure drop change.

ＩＴＥＲ‐ＣＳモデル・コイルとＣＳインサート・コイルの実験結果 ＣＳモデル・コイルの圧力損失及び流量配分

International Thermonuclear Experimental Reactor (ITER) Engineering Design Activities (EDA) have been performed in collaboration with Japan, the European Union, the Russian Federation, and the United States of America. In ITER EDA, the Central Solenoid Model Coil (CSMC) and CS insert coil have been fabricated to demonstrate the realization of ITER superconducting magnet. The conductor used in the CSMC and CS insert coil is a forced flow cable-in-conduit conductor (CICC) cooled by supercritical helium. For CICC, a pressure drop characteristic is one item of important information from the viewpoint of understanding of heat transfer in the conductor and thermodynamics design of a cryogenic system with a helium circulation pump for coil cooling. The ITER conductor has a central channel made of a spiral tube of thin stainless steel tape and a spring tube made of INCONEL in the bundle having about 1, 100 superconducting strands. The helium gas flow into the bundle and central channel in parallel and the pressure drop characteristic are determined from flow balance between the bundle region and the central channel because of different flow friction factor characteristics of both. In the past, pressure drop measurement for various bundle types of CICC composed of a few hundreds strands, such as Demo Poloidal Coil (DPC), has been well performed at 4-K supercritical helium. Recently, the friction factors of the bundle region and the central channel of the ITER conductor have been measured at room temperature. The pressure drop characteristics of an ITER relevant sub size conductor, such as QUELL have been measured at 4K. However, there are no measurement results in the practical coil operation conditions (4-K supercritical helium condition) for an ITER full-size conductor. In a CSMC experiment, the pressure drop characteristic of ITER full size conductor has been measured in a practical operation condition for the first time at an ITER CS model coil test facility in the Naka Fusion Research Establishment, Japan Atomic Energy Research Institute (JAERI).

Stability under transverse field pulse of the Nb/sub 3/Sn ITER cable-in-conduit conductor

A section of the ITER (International Thermonuclear Experimental Reactor) conductor, left from the winding of the Central Solenoid Model Coil, was tested for stability under pulsed transverse field in the SULTAN facility, at background field up to 11 T and dc current up to 100 kA. The test was repeated at a broad range of operating temperature and for different current distribution in the strand bundle. The ability to withstand transverse field transients is compared with the predicted energy margin. For large current unbalance, the degradation of transient stability is well correlated with the decrease of quench temperature in dc mode. The crucial role of the current sharing among subcables for the stability of the cable-in-conduit conductors is assessed for short samples as well as for the ITER coils. Basic recommendations are drawn for the improvement of the conductor design.

Stability and quench behavior for pulsed tests on ENEA 12 T Nb3Sn CICC coil

The Central Solenoid (CS) of ITER (International Thermonuclear Experimental Reactor) is designed to operate at 13 T, 0.5-1 T/s, 5.8 K. There is an experimental evidence that reduced performances in variable field, not explained by AC losses, could occur in CIC (cable-in-conduit) conductors. Such ramp rate limitations have to be well understood in order to be avoided in the ITER conductors. In the ENEA Frascati Laboratory a facility has been assembled to test in pulsed regimes (up to 8T, 4 T/s, 12 K) the ENEA Nb3Sn CICC coil. The characteristics of the coil (dimensions, CIC conductor, strand designed for use in variable field). Main objectives of the program are experimental data on ramp rate limitations, pulsed-current distribution in the cable, limiting currents and global AC losses. This paper reports the results of the measurements carried out up now, and those of the simulations performed by means of the thermo-hydraulic analysis code SARUMAN.

Design and R&amp;D results of the joints for the ITER conductor

In the scope of the Engineering Design Activity (EDA) for ITER (International Thermonuclear Experimental Reactor), a R&D task has been launched in 1993 for development, manufacturing and testing of subsize and full size joints, to be used in the layer-to-layer and pancake-to-pancake connections of the ITER coils. A set of design criteria and manufacturing choices is initially agreed as a common basis for the work carried out in parallel by the four ITER parties (Japan, European Union, United States, Russian Federation). The staged work includes variations of layout parameters in subsize samples, technological investigations for manufacturing methods and performance analysis. As a last stage of the task, full size joint samples, manufactured from sections of the ITER conductors, are tested in the PTF and SULTAN test facilities under ITER relevant operating conditions.

DC resistance and eddy current losses in the ITER conductor joint

The central solenoid (CS) and the toroidal field (TF) coils of the International Thermonuclear Experimental Reactor (ITER) are wound from superconducting (SC) Nb/sub 3/Sn Cable-in-Conduit-Conductor (CICC). Electrical joints are an important aspect of the coil design, since local heating from resistive losses can drive the superconductor into a less stable regime and possibly initiate a quench. A good joint design will satisfy the broadest requirement which is to minimize the power dissipation during normal coil operations. This can be achieved by satisfying two, more specific, and in some ways, conflicting requirements: (1) to provide a low DC resistance, and (2) to minimize the effects of Joule heating from eddy currents induced by transient magnetic fields. This paper presents a finite element analysis of the latest conductor joint concept and evaluates its electromagnetic characteristics with respect to these two requirements. The analysis extends the usual practice of representing the presence of superconductors with surface conditions by explicitly modeling the last stage of the twisted cable. Although certain modeling approximations limit the accuracy of the results, some valuable insights are gained, such as which approximations cannot be made, and the effects of various design details on the DC joint resistance and Joule heating. >

Effects of diffusion barrier on the magnetization cycles of Nb/sub 3/Sn strand for ITER conductor

High performance Nb/sub 3/Sn strands for the high field ITER (International Thermonuclear Experimental Reactor) magnets have been manufactured using diffusion barriers among the multifilamentary cores and the stabilizing copper. In the conductor development programme, billets using a single Ta or a double NbV barrier have been assembled. The barrier is shown to have an influence on magnetization characteristics. The measurements on samples from the two types of strands are shown and analyzed to correlate them with the magnetic behaviour of Nb at low field. The results show that at low field hysteretic losses are greatly increased by the pure Nb of the barrier. At high field, losses of samples with Ta barrier are larger, due to their higher J/sub c/.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Stability estimations and quench simulations for the ITER conductors

Quench simulations and stability estimations for the International Thermonuclear Experimental Reactor (ITER) are discussed. Especially numerical issues and associated benchmark actions are summarized. Satisfactory agreement between the various codes from the 4 ITER parties is now obtained after numerical convergence problems have been resolved. However, these require confirmation by experiments on relevant conductor geometries. In multistage cables, a non-uniform current distribution within the cable affects the stability of the conductor. A possible mechanism for the non-uniform distribution is flux loops between the strands or cable substages as the current is ramped up or down. A preliminary estimation of stability with non-uniform current distribution is also discussed.