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Fusion Engineering and Design 01/2011; 86(6):1418-1421. · 1.49 Impact Factor
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ABSTRACT: The quench detection system in ITER superconducting magnet systems is a real challenge for the poloidal field (PF) system and the central solenoid (CS) due to their fast-pulsed operation. An existing CEA code allowing very precise magnetic field calculations in tokamaks (Traps) has been upgraded (TrapsAv) to calculate the inductive voltages induced across any coil sub-element during a plasma scenario. The envisaged primary detection system is based on voltage measurements. For the CS, a detection system is proposed and discussed based on the central difference average balance of 3 neighboring double pancakes. For the TF system, the presently considered compensation circuit in ITER is a strip co-wound on to the conductor. A simple analytical calculation is proposed to estimate the voltage during a plasma disruption or during the scenario across the co-wound strip and across the conductor as well, due to the toroidal component of the plasma current.
IEEE Transactions on Appiled Superconductivity 07/2010; · 1.04 Impact Factor
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ABSTRACT: The JT-60SA Toroidal Field coil design has been modified resulting in considerable material savings. The casing cooling loop is important during the beginning of cooldown. This study models the cooling phase foreseen for coil testing and coil operation. Data and experience gathered during the extensive cooling tests performed on 70 large W7-X coils at the CEA (Saclay) cryomagnetic test facility provide the basis knowledge to model and forecast the thermal behavior of other large magnets for thermonuclear fusion. Using coil material thermal properties, refrigeration limitations, procedure and control, a simple model of JT-60SA TF coil winding and casing cooling is proposed. Thermal gradients inside the coil are evaluated and discussed. The calculated cooldown time and the required cryogenic power will be used as design input for the coil testing facility.
IEEE Transactions on Appiled Superconductivity 07/2010; · 1.04 Impact Factor
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ABSTRACT: Recent progresses in Bi2212 wires have proved its suitability for round wire developments and high field magnet development. High energy storage magnets can be foreseen leading to cable developments. In order to prepare such work CEA Saclay has developed in collaboration with the Nexans company a Bi 2212 wire having 18 sub-elements. The round wire has been twisted to study losses and twisting degradation. The results are presented here and compared to losses measurements made on already existing ribbons.
IEEE Transactions on Appiled Superconductivity 07/2009; · 1.04 Impact Factor
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A. Pizzuto,
L. Semeraro,
L. Zani,
P. Bayetti,
A. Cucchiaro,
P. Decool,
A. della Corte,
A. Di Zenobio,
N. Dolgetta, J.L. Duchateau, [......],
B. Lacroix,
L. Muzzi,
S. Nicollet,
G.M. Polli,
C. Portafaix,
L. Reccia,
S. Turtu,
J.-M. Verger,
R. Villari,
K. Yoshida
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ABSTRACT: The broader approach agreement between Europe and Japan includes the construction of a fully superconducting tokamak, the JT-60 Super Advanced (JT-60SA), as a satellite experiment to ITER. In particular, the whole Toroidal Field magnet system, described in this paper, will be provided to Japan by the EU. All the TF coil main constituents, i.e. conductor, winding pack, joints, casing, current leads, are here presented and discussed as well as the design criteria adopted to fulfil the machine requirements. The results of the analyses performed by the EU and JA to define and assess the TF magnet system conceptual design are reported and commented. Future work plan is also discussed.
IEEE Transactions on Appiled Superconductivity 07/2008; · 1.04 Impact Factor
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L. Zani,
A. Pizzuto,
L. Semeraro,
D. Ciazynski,
A. Cucchiaro,
P. Decool,
A. della Corte,
A. Di Zenobio,
N. Dolgetta, J.-L. Duchateau, [......],
S. Nicollet,
L. Petrizzi,
C. Portafaix,
G. Ramogida,
S. Roccella,
B. Turck,
S. Turtu,
J.-M. Verger,
R. Villari,
K. Yoshida
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ABSTRACT: The upgrade of JT-60U to JT-60 Super Advanced (JT-60SA), a fully superconducting tokamak, will be performed in the framework of the Broader Approach (BA) agreement between Europe (EU) and Japan. In particular, the Toroidal Field (TF) system, which includes 18 coils, is foreseen to be procured by France, Italy and Germany. This work covers activities from design and manufacturing to shipping to Japan. The present paper is mainly devoted to the analyses that lead to the conductor design and to the technical specifications of the joints for the JT-60SA TF coils. The conductor geometry is described, which is derived from Cable-In-Conduit concept and adapted to the actual JT-60SA tokamak operating conditions, principally the ITER-like scenario. The reported simulations and calculations are particularly dealing with the stability analysis and the power deposition during normal and off-normal conditions (AC losses, nuclear heating). The final conductor solution was selected through a trade-off between scientific approach and industrial technical orientation. Besides, the TF system connections layout is shown, derived from the industrially assessed twin-box concept, together with the associated thermo-hydraulic calculations ensuring a proper temperature margin.
IEEE Transactions on Appiled Superconductivity 07/2008; · 1.04 Impact Factor
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K. Yoshida,
K. Kizu,
K. Tsuchiya,
H. Tamai,
M. Matsukawa,
M. Kikuchi,
A. della Corte,
L. Muzzi,
S. Turtu,
A. Di Zenobio,
A. Pizzuto,
C. Portafaix,
S. Nicollet,
B. Lacroix,
P. Decool, J.-L. Duchateau,
L. Zani
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ABSTRACT: The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.
IEEE Transactions on Appiled Superconductivity 07/2008; · 1.04 Impact Factor
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K. Kizu,
K. Tsuchiya,
K. Yoshida,
M. Edaya,
T. Ichige,
H. Tamai,
M. Matsukawa,
A. della Corte,
A. Di Zenobio,
L. Muzzi,
S. Turtu, J.L. Duchateau,
L. Zani
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ABSTRACT: The conductor for central solenoid (CS) and equilibrium field (EF) coils of JT-60 Super Advanced (JT-60SA) were designed. The conductor for CS is Nb<sub>3</sub>Sn Cable-In-Conduit (CIC) conductor with JK2LB jacket. EF coil conductors are NbTi CIC conductor with SS316LN jacket. The field change rate (3.9 T/s), faster than ITER generates the large AC loss in conductor. The analyses of current sharing temperature (T<sub>cs</sub>)margins for these coils were performed by the one-dimensional fluid analysis code with transient heat loads. The margins of these coils are 1 K for the plasma standard and disruption scenarios. The minimum T<sub>cs</sub> margin of CS conductor is 1.2 K at plasma break down (BD). The margin is increased by decreasing the rate of initial magnetization. It is found that the disruption mainly impacts the outer low field EF coil. The disruption decreases the T<sub>cs</sub> margin of the coil by >1 K. A coupling time constant of <100 ms, Ni plating, and a central spiral are required for NbTi conductor.
IEEE Transactions on Appiled Superconductivity 07/2008; · 1.04 Impact Factor
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ABSTRACT: The reflexion about DEMO, a fusion demonstration reactor, has already been initiated in the framework of European tasks and the decision to build ITER at CEA Cadarache has even reinforced the interest for such studies. A key component of this fusion demonstration reactor is the superconducting magnetic system, which represents in ITER 30% of the investment cost. Which superconducting materials are adequate for DEMO and are the emerging HTc superconducting material a necessary technology which cannot be avoided? This crucial question will be examined and discussed, taking into account key parameters of the project such as: - the toroidal magnetic field Bt which plays a determining role in the fusion power and the amplification factor, in association with the major radius R of the machine which is a characteristics of the reactor size. - the operating temperature of the magnet system which can affect the global efficiency of the reactor. As a matter of fact the electrical power associated with the cryogenic refrigeration of the magnet will be estimated and discussed. Two temperatures will be considered: 5 K and 20K. The dimensions of the TF magnet system can be estimated, thanks to a simplified approach. These dimensions have an impact on the machine and enable to obtain a realistic integration of the superconducting magnet system with all aspects of the machine within a system approach.
Journal of Physics Conference Series 03/2008; 97(1):012038.
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ABSTRACT: Within the ITER Poloidal Field conductor design validation, the Poloidal Field Conductor Insert (PFCI) has been manufactured and will be tested in the Central Solenoid Model Coil (CSMC) facility at JAEA Naka (Japan). In this test facility, the PFCI can be tested under ITER-relevant operating conditions, the field produced by the CSMC being varied to simulate the real situation of the PF coils in ITER. Predictive analyses have been performed in order to study the electromagnetic and thermal-hydraulic behaviour of the PFCI, under two scenarios proposed for pulsed current tests. During these scenarios, simulations have been performed with the THEA code, in which classical formulas for the AC losses in a cable have been introduced. The study focuses on the lower part of the winding, which is a 44 m long conductor including a joint. It covers the sample thermal-hydraulic behaviour with particular emphasis on the losses. Due to the overcompaction in the joint area, the total energy dissipated during a scenario can be equivalent in the joint and in the conductor, in spite of the reduced length of the joint (0.45 m). This particular point is discussed and has led to the analysis of the temperature margin in the joint.
Journal of Physics Conference Series 03/2008; 97(1):012201.
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ABSTRACT: In parallel to ITER preparation, power plant conceptual studies have been carried out in Europe as a guideline for the main design parameters of a fusion demonstration reactor, the construction of which is scheduled to start twenty years hence. A key component of this fusion demonstration reactor is the superconducting magnetic system, which represents in ITER 30% of the investment cost. A high toroidal confinement field B<sub>t</sub>, can help to moderate the major radius of the tokamak, and thereby the investment cost of the machine. Two versions of such a demonstration reactor, with different values of B<sub>t</sub> will be compared in terms of volume, radial extension for the toroidal field magnet inner leg, central solenoid available flux swing and investment cost.
IEEE Transactions on Appiled Superconductivity 07/2007; · 1.04 Impact Factor
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ABSTRACT: In the cable-in-conduit-conductor (CICC) design of the toroidal field system for the international thermonuclear reactor (ITER) Nb<sub>3</sub>Sn is used as superconductor material. Considering the single strand performance, the crucial characteristic is the strain dependence of the critical current. Within this context, the performance of the CICC under strain is determined by the behaviour of the single strands and additional effects related to the manufacturing process. In the framework of the European fusion technology program a task has been started to investigate single strands as well as sub-size CICC performance using different cable layouts (9, 45 and 180 strands). For this systematic approach, parameters such as the void fraction, the number of pure copper strands, the void fraction or the cabling pattern have been varied. To examine the critical properties in detail, the available test facility, consisting of two experimental setups, is capable to measure the strain dependence in magnetic fields up to 14 T at 4.2 K, by applying an axial load to the samples. Measurements on such sub-size CICC samples are presented and compared to the expected performance.
IEEE Transactions on Appiled Superconductivity 07/2007; · 1.04 Impact Factor
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ABSTRACT: The refrigeration power associated with the superconducting magnets and cryopumps of a steady-state fusion reactor is not negligible. The power has to be minimized because it plays a role in the power station global efficiency and in the required amplification factor Q. On the one hand, the long plasma discharges obtained in December 2003 on TORE SUPRA give an insight of the cryogenic losses that might be expected for a steady-state fusion reactor equipped with superconducting magnets. The superfluid bath of the windings in TORE SUPRA allows a simple calorimetric estimation of the cryogenic losses through the temperature evolution of the bath during the long discharge. The different kinds of losses in TORE SUPRA are estimated, discussed and explained. Not all of them will be present in a real reactor. On the other hand, in the framework of ITER preparation, the magnet system and the associated refrigerator have been dimensioned taking into account again all kinds of cold losses. This exercise is important because ITER, by its size, could be relevant to the steady-state reactor situation regarding refrigeration. Based on TORE SUPRA experiment and ITER design it is, therefore, possible to propose for the first time a preliminary figure for the cryoplant power of a steady-state reactor. The order of magnitude of the cryoplant power is ten times lower than that of the fusion reactor recycled power which can be considered acceptable.
Nuclear Fusion 02/2006; 46(3):S94. · 4.09 Impact Factor
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ABSTRACT: The ITER Toroidal Field (TF) magnet is made of 18 coils. The 7 winding‐pack modules constituting their structure, contain each one double pancake Cable‐In‐Conduit‐Conductor (CICC) inserted in a stainless steel radial plate. The winding pack is enclosed inside a thick stainless steel case. During a safety discharge of the TF system, eddy currents and associated heat generation are induced in the plates and in the case. Such fast discharges have been performed during Phase I and Phase II experiment of the Toroidal Field Model Coil (TFMC) tests to investigate heat generation and transfer. Simple models have been developed to predict the behaviour of the coil during these transients. These models have been validated by calorimetric measurements. These models can also be used to estimate the temperature increase of the TF Conductor during a plasma disruption and to investigate whether a quench can happen. © 2004 American Institute of Physics
AIP Conference Proceedings. 06/2004; 710(1):693-700.
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ABSTRACT: In a phase II experiment on the International Thermonuclear Experimental Reactor (ITER) toroidal field model coil (TFMC) the operation limits of its 80 kA Nb3Sn conductor were explored. To increase the magnetic field on the conductor, the TFMC was tested in the presence of another large coil: the Euratom LCT coil. Under these conditions the maximum field reached on the conductor was around 10 T. This exploration has been performed at constant current, by progressively increasing the coil temperature and monitoring the coil voltage drop in the current sharing regime. Such an operation was made possible thanks to the very high stability of the conductor. The aim of these tests was to compare the critical properties of the conductor with expectations and to assess the ITER TF conductor design. These expectations are based on the documented critical field and temperature dependent properties of the 720 superconducting strands which compose the conductor. In addition the conductor properties are highly dependent on the strain, due to the compression appearing on Nb3Sn during the heat treatment of the pancakes and related to the difference in thermal compression between Nb3Sn and the stainless steel jacket. No precise model exists to predict this strain, which is therefore the main information which is expected from these tests. The method to deduce this strain from the different tests is presented, including a thermohydraulic analysis to identify the temperature of the critical point and a careful estimation of the field map across the conductor. The measured strain has been estimated in the range −0.75% to −0.79%. This information will be taken into account for ITER design and some adjustment of the ITER conductor design is under examination.
Superconductor Science and Technology 03/2004; 17(5):S241. · 2.66 Impact Factor
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ABSTRACT: The construction and testing of the Toroidal Field Model Coil (TFMC) is part of one of the ITER large R&D projects. The main goal was to demonstrate the feasibility and the mechanical integrity of the design. One of the highlights of the first test phase was to measure the current sharing temperature, T<sub>CS</sub>, of the conductor by heating the helium entering from the inlet. Because neither temperature sensors nor voltage taps are positioned inside the coil, only the helium inlet temperature and the voltage along the whole conductor length can be used for the evaluation of T<sub>CS</sub>. In addition, an inner pancake joint is located at the inlet in a rather high magnetic field and the peak field region is only about 1.5 m apart from the joint. The determination of the T<sub>CS</sub> relies on the exact knowledge of the thermohydraulics of both the joint and the conductor region. The paper describes and compares the different numerical models used for the evaluation of the T<sub>CS</sub>. Nine T<sub>CS</sub> tests at different coil currents were performed, all ending in a quench. The measured T<sub>CS</sub> is in good agreement with the expectations.
IEEE Transactions on Appiled Superconductivity 07/2003; · 1.04 Impact Factor
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ABSTRACT: A new kind of superconductor, using the cable-in-conduit concept, is emerging, mainly in the context of fusion activity. At present no large Nb3Sn magnet in the world is operating using this concept. The difficulty of this technology, which has now been studied for 20 years, is that it requires major advances in several interconnected new fields, such as handling a large number (1000) of superconducting strands, high current conductors (50 kA), forced flow cryogenics, Nb3Sn technology, low loss conductors in pulsed operation, high current connections and high voltage insulation (10 kV), as well as demonstration of its economical and industrial feasibility. CEA has been very much involved, during the past ten years, in this development, which took place in the framework of the NET and ITER technological programmes. One major milestone was reached in 1998-1999 with the successful tests by Euratom-CEA of three full size conductor and connection samples in the SULTAN facility in Switzerland.
Nuclear Fusion 05/2002; 41(2):223. · 4.09 Impact Factor
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ABSTRACT: In the framework of the preparation for the realization of the international thermonuclear experimental reactor (ITER), the construction and test of relevant models of seven different parts of the reactor was decided. Two of them are related to the superconducting coils: the toroidal field model coil (TFMC) and the central solenoid model coil (CSMC). For these superconducting coils, due to the expected high values of the current (≥60 kA) and voltage (≥5 kV with respect to the ground) the adopted technology was that of cable in conduit conductor (CICC). Until recently, little experience of this technology existed. Therefore, an extensive research and development programme has been carried out, in the last 10 years, by the ITER partners and particularly in Europe, to design, industrialize and test these large conductors and their joints. The EURATOM associations CEA and ENEA played a leading part in this phase. The CICC concept is described and the results of the developments are presented. About 7 km of conductors were manufactured in the industry and for that more than 10 tonnes of Nb3Sn strands were produced in Europe. In this large programme, Europe is particularly in charge of the TFMC, which will be tested this summer at Forschung Zentrum Karlsruhe (Germany). In the framework of this programme, three full size conductors and joint samples were tested at the European Sultan test facility (Centre de Recherches de Physique des Plasmas, Villigen, Switzerland), to validate the technological choices and check that the ITER specifications were met. The results of these tests are presented in detail. Starting from the strand critical properties, the conductors made of about 1000 strands did reach their expected performance. The joints of these large conductors are very special and delicate components. Their behaviour was quite successful and the joint resistance of these samples (of the order of 1 nΩ) was well within the specifications.
Superconductor Science and Technology 04/2002; 15(6):R17. · 2.66 Impact Factor
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ABSTRACT: The test of the Toroidal Field (TF) Model Coil of the International Thermonuclear Experimental Reactor (ITER) has been the opportunity to measure the DC resistances of all the joints of a real coil and to compare them to values previously measured on prototype full-size joint samples. This paper describes and discusses the different methods used for measuring all the joint resistances, and gives the results of joint resistance measurements (1-2 nΩ range) as function of the coil current up to the maximum value of 80 kA. Comparisons with resistances measured on prototype joints in relevant field/current conditions are presented and discussed.
IEEE Transactions on Appiled Superconductivity 04/2002; · 1.04 Impact Factor
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ABSTRACT: During the last phase of the EDA (Engineering Design Activities) of the ITER (International Thermonuclear Experimental Reactor) project, a fully superconducting PF (Poloidal Field) system has been designed to cope with the requirements of the FEAT (Fusion Energy Advanced Tokamak) version of ITER: providing the magnetic field required to develop, shape and control the 15 MA plasma during the 900 s of a typical plasma scenario. The CS (Central Solenoid), divided into 6 Nb<sub>3</sub>Sn subcoils, and the six outer NbTi PF coils will experience severe heat loads especially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation, and ac losses (30 to 300 kJ in the PF coils). The ac losses along the PF coil pancakes are deduced from accurate magnetic field computations performed by the TRAPS code (analytical integration of the Biot-Savart law over the cross section of 3-D current elements). Using as input these heat loads, including thermal radiation and nuclear heating, a thermal-hydraulic analysis of the PF coil cable-in-conduit conductor is performed with a finite element code, GANDALF: the temperature increases (0.1 to 0.3 K) are computed, the temperature margins of the conductor are thus evaluated.
IEEE Transactions on Appiled Superconductivity 04/2002; · 1.04 Impact Factor
