Joint resistance measurements of pancake and terminal joints for JT-60SA EF coils

To evaluate the joint fabrication technology for the JT-60SA EF coils, joint resistance measurements were conducted using a sample consisting of pancake and terminal joints. Both joints are shake-hands lap joints composed of cable-in-conduit conductors and a pure copper saddle-shaped spacer. The measurements demonstrated that both joints fulfilled the design requirement. Considering these measurements, the characteristics of both joints were investigated using analytical models that represent the joints. The analyses indicated that the characteristics of the conductors used in the joints affect the characteristics of the joints.     

CEA studies and qualifications prior to the JT-60SA TF coils manufacture

Following the first conceptual design activity in which the general design of the JT-60SA TF system was defined in agreement with all the participants in the project (CEA, ENEA, F4E), a second phase dealing with the detailed design was engaged by each of the voluntary contributors. For the French part which includes the procurement of 9 of the TF winding packs and their integration in the casing, an industrial contract was signed mid 2011 with Alstom (France). Several actions have been carried out to prepare the manufacturing phase.To precisely define one of the main interfaces which is the temporary electrical connection of the coils to the current leads during cold test in the CEA facility as well as their final connection to the feeders at the Naka site, a design compatible with both requirements was developed by CEA, supported by the previous developments led on the joints and assembly techniques.In addition to prepare the coils manufacture, hydraulic qualification was led on the first conductor qualification length to set the parameters which will be used by the coils manufacturer for conductor acceptance.At last, mechanical characterizations of both the conductor and of the empty compacted jacket were performed in order to define as precisely as possible the elastic and plastic properties of these components. These are crucial properties used during the bending process which is one of the most important operations during the winding pack manufacture. These data will be very helpful for the winding machine parameters settings as well as for designing the local bending tooling needed to shape the conductors extremities at the connection area and at the double pancakes joggles.     

Fabrication and installation of equilibrium field coils for the JT-60SA

Recently, fabrication of the first superconducting coil in JT-60SA tokamak (EF4 coil) was finished. EF4 coil has ten double pancake (DP) coils. All DP coils were stacked up to form the winding pack. In order to check the manufacturing error of DP coils, their circularities (in-plane ellipticities) were evaluated for all DP coils. Positions of conductors for each DP coil were measured before curing process. Error bars of the current centers, which were used for the index for DP coil's circularity, ranged between 1.1 and 2.5 mm. During stacking the DP coils, the positions of these coils were optimized in order to cancel the error of circulation of the winding pack. As the result, the manufacturing error of the radial current center was achieved 0.6 mm for the winding pack. This value was an order of magnitude smaller than the required manufacturing error of EF4 coil.     

A coil test facility for the cryogenic tests of the JT-60SA TF coils

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy (“Fusion for Energy”, F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature.The project is presently in the detailed design phase.     

2D thermal analysis for heat transfer from casing to winding pack in JT-60SA TF coils

In the framework of the Broader Approach Agreement, Europe is involved in the design activities for the Japanese Tokamak JT-60SA, investigating, among several issues, the operation of the superconducting TF magnets and their subsystems, aimed at the reactor conceptual design definition. In particular, one of the main critical aspects to study is the heating of the conductor due to both direct component of energy deposited by neutrons and by secondary gamma generated during plasma operation, and heat generated by the radiation on casing and transferred to the winding pack. Indeed, the operating temperature and the relevant temperature margin (i.e. the operating safety margin) of the magnet will depend strongly on the heat loads and on the capability of the coolant to remove it. Furthermore, the heat power to the conductor will depend on several aspects, namely the thickness of insulating material, the mass flow rate of helium flowing in the conductors and its thermodynamic properties at operating conditions, and the layout of the superconductors constituting the winding. Moreover, a crucial aspect in the final design will be the presence and position of the casing cooling channels. In this paper a 2D sensitivity analysis of heat transfer from casing to winding pack with respect to cooling channels number and position is presented, based on the reference layout of the magnet. As a result, we evaluated the optimum number and positioning of cooling channels needed, as a trade-off between magnet operating limits and available cryogenic power and if, at limit, they could be even neglected in normal operation, keeping dwell-time within reasonable values.     

Infrared thermography inspection for monoblock divertor target in JT-60SA

Carbon fiber composite (CFC) monoblock divertor target is required for power handling in JT-60SA. Quality of the targets depends on a joining technology in manufacturing process. To inspect the quality of more than 900 target pieces, efficient non-destructive inspection is needed. An infrared thermography inspection (IR inspection), has been proposed by ITER and IRFM, where the quality between CFC and a cooling tube is examined by a use of transient thermal response at a rapid switch from hot to cold water flow. In JT-60SA divertor target, a screw tube will be employed to obtain high heat transfer efficiency with simple structure. Since the time response of the screw tube is much faster than that of smooth tube, it is required to confirm the feasibility of this IR inspection. Thus, the effect of joining defects on transient thermal response of the targets has been investigated experimentally by using the mock-up targets containing defects which are artificially made. It was found that the IR inspection can detect the defects. Moreover, screening criteria of IR inspection for acceptable monoblock target is discussed.     

Upgrading the NIFS superconductor test facility for JT-60SA cable-in-conduit conductors

The superconductor test facility at the National Institute for Fusion Science (NIFS) was upgraded to test cable-in-conduit (CIC) conductors for the JT-60SA equilibrium field (EF) coil. Supercritical helium (SHe) lines were assembled with transfer tubes and a heat exchanger. The CIC conductor was covered with a thermal insulation vessel, filled with gas helium at atmospheric pressure. The temperature of the conductor was varied using a film heater attached to an inlet pipe. Critical current (Ic) and current sharing temperature (Tcs) measurements of the prototype CIC conductor were carried out successfully in the upgraded test facility. During the measurements, the conductor temperature was precisely controlled.     

Quench detection of fast plasma events for the JT-60SA central solenoid

The JT-60 is planned to be modified to a full-superconducting tokamak referred to as the JT-60 Super Advanced (JT-60SA). The maximum temperature of the magnet during its quench might reach the temperature of higher than several hundreds Kelvin that will damage the superconducting magnet itself. The high precision quench detection system, therefore, is one of the key technologies in the superconducting magnet protection system.The pick-up coil method, which is using voltage taps to detect the normal voltage, is used for the quench detection of the JT-60SA superconducting magnet system. The disk-shaped pick-up coils are inserted in the central solenoid (CS) module to compensate the inductive voltage. In the previous study, the quench detection system requires a large number of pick-up coils. The reliability of quench detection system would be higher by simplifying the detection system such as reducing the number of pick-up coils. Simplifying the quench detection system is also important to reduce the total cost of the protection system. Hence the design method is improved by increasing optimizing parameters. The improved design method can reduce the number of pick-up coils without reducing the sensitivity of detection; consequently the protection system can be designed with higher reliability and lower cost. The applicability of the disk-shaped pick-up coil for quench detection system is evaluated by the two dimensional analysis. In the previous study, however, the analysis model only took into account the CS, EF (equilibrium field) coils and plasma. Therefore, applicability of the disk-shaped pick-up coil for the quench detection system remains open question because the fast plasma events, such as disruption, mini collapse and ELM (edge localized mode), directly influences on the voltage of pick-up coil making the quench signal undetectable. Consequently, a new analysis model proposed in the present paper was designed to avoid this difficulty by introducing the passive coil series such as vacuum vessel and stabilizer. The influence of fast plasma events is absorbed by passive coil series like real system, and the evaluation of applicability can be examined in detail. The analysis results show that the disk-shaped pick-up coil is applicable whenever the standard operation, disruption, mini collapse and ELM.     

Magnet design of toroidal field coils for the UWMAK-Tokamak systems

The design of toroidal field coils for the UWMAK series of Tokamak reactor designs is described. These are cryogenically stable coils cooled in liquid helium to 4.2 K. Each individual turn of composite conductor of TiNb plus matrix conductor is epoxied into a groove in a thin disk structure. The magnet is divided into 12, 18 or 24 sectors; each sector is comprised of 15–20 thin disks which are spaced and bolted together to form a rigid structure with all disk surfaces exposed to cooling. The overall shape of each ‘D’ magnet sector is chosen so that only constant tension forces are present. Bending forces do occur but only near transition sections from the D to the central straight section of each coil. This method of rigid mounting should be compared with loose ‘jelly-roll’ windings on a central coil form, a more typical magnet fabrication technique. The design procedure is for the composite conductor TiNb plus copper (or aluminum) to be mounted in stainless steel (or aluminum alloy) disks. Full stability is obtained for strains less than 0.2% for steel support and less than 0.4% for aluminum supports based on stress-strain resistivity experiments in progress. The use of high purity aluminum conductor and high strength aluminum alloy structure reduces costs significantly dependent only on the orderly development of new aluminum TiNb composite conductors. The amount of TiNb is conservatively chosen to carry full current at 5.2 K although operation at 4.2 K is planned and full recovery to the superconducting state could be obtained with full current wire quantities selected at 4.3 K. This conservative choice doubles the amount of TiNb used at 8 tesla but provides an extra temperature rise of ΔT = 0.9 K above expected usual temperature excursions.Magnet safety and protection is based on the natural mutual coupling of many coils which are closely coupled to each other. If one coil loses current, the other coils increase their currents to keep the flux as constant as possible. The uncoupled flux and companion field energy would be discharged by a high voltage power supply temporarily set to discharge the one bad coil. Such sub-division and partial energy removal requires that there be substantial subdivision of coils into many separate dewars, so that problems can be isolated. An expression for the magnetic forces on sectioned toroidal field coils is given in closed form and is used to compute the shape of a specific coil. Data obtained here are shown to be in good agreement with those given by more complex procedures. The most severe structural design requirement is based on simultaneous loss of current in two adjacent sectors. The remaining sectors attempt to straighten out into a solenoid which compresses the structure between coils except beside the bad coil or coils where tension might exist. Such current loss in two adjacent sectors is considered an extremely unlikely occurrence since the discharge procedure mentioned above takes place in less than 1 min so that simultaneous refers to a 1 min overlap. Because of such rapid amelioration of the causes of current change and flux motion, no temperatures can exceed room temperature during the orderly shutdown of one or two coils. In general, the study illustrates that fully stable magnets using composite conductors should be engineered without major uncertainties according to straightforward scientific concepts. While subsequent designs will undoubtedly include improvements there is no reason to expect that superconductivity implies venturesome unknown TF coil performance.     

Fundamental welding R&D results for manufacturing vacuum vessel of JT-60SA

The real vacuum vessel (VV) manufacturing of JT-60SA has started since November 2009 at Toshiba. Prior to starting manufacturing, fundamental welding R&Ds had been performed by three stages. In the first stage, primary tests for screening welding method were performed. In the second stage, the trial welding for 1 m-long straight and curved double shell samples were conducted. The dependences of welding quality and distortion on the welding conditions, such as arc voltage and current, setting accuracy, welding sequence, and the shape of grooves were studied. In addition, welding condition with low heat input was explored. In the last stage, fabrication sequence was confirmed and established by the trial manufacturing of the 20° upper half mock-up [1]. This paper presents the R&D results obtained in the first and second stages.     

Progress in development and design of the neutral beam injector for JT-60SA

Neutral beam (NB) injectors for JT-60 Super Advanced (JT-60SA) have been designed and developed. Twelve positive-ion-based and one negative-ion-based NB injectors are allocated to inject 30 MW D⁰ beams in total for 100 s. Each of the positive-ion-based NB injector is designed to inject 1.7 MW for 100 s at 85 keV. A part of the power supplies and magnetic shield utilized on JT-60U are upgraded and reused on JT-60SA. To realize the negative-ion-based NB injector for JT-60SA where the injection of 500 keV, 10 MW D⁰ beams for 100 s is required, R&Ds of the negative ion source have been carried out. High-energy negative ion beams of 490–500 keV have been successfully produced at a beam current of 1–2.8 A through 20% of the total ion extraction area, by improving voltage holding capability of the ion source. This is the first demonstration of a high-current negative ion acceleration of >1 A to 500 keV. The design of the power supplies and the beamline is also in progress. The procurement of the acceleration power supply starts in 2010.     

Development of Negative-Ion Based NBI System for JT-60

The negative-ion based neutral beam injector (N-NBI) for JT-60 has been developed for plasma core heating and neutral beam current drive in higher density plasmas. Construction of the N-NBI system was completed in 1996, and just after this completion, the efforts to increase beam power and beam energy started. The N-NBI system has already operated with negative ion beams with 14.3 A at 380 keV of deuterium and with 18.5A at 360 keV of hydrogen. During N-NBI experiments on JT-60, a deuterium neutral beam power of 5.2MW at 350keV has been injected for 0.7s stably, and the response of the JT-60 plasma to high energy beam injection with the N-NBI has been confirmed to be in agreement with a theoretical prediction.     

Status of European manufacture of Toroidal Field conductor and strand for JT-60SA project

In the framework of the JT-60SA project, part of the Broader Approach (BA) agreement, EURATOM provides to Japan, the Toroidal Field (TF) magnet system, consisting of 18 superconducting coils. The procurement of the conductor for the TF coils is managed by Fusion for Energy, acting as EU representative in the BA agreement. The TF conductor procurement is split into two contracts, one dedicated to the production of Niobium Titanium (NbTi) and Cu strand and the other to TF conductor production through strand cabling and cable jacketing operations.The TF conductor is a rectangular-shaped cable-in-conduit conductor formed by 486 (0.81 mm diameter) strands (2/3 NbTi–1/3 Cu) wrapped in a stainless steel foil and embedded into a stainless steel jacket.The 18 TF coils require (including spares) 115 ‘Unit Lengths’ (UL) of such conductor, each 240 m long for a total of about 28 km. Correspondingly about 10,000 km for NbTi and 5000 km for Cu strand are produced.The Japanese company Furukawa Electric Co. (FEC) is in charge of TF strand manufacture while the Italian company Italian Consortium for Applied Superconductivity (ICAS) is in charge of cabling and jacketing of TF conductor ULs. In the paper, we provide information on the production stages presently achieved in TF strand and conductor contracts.     

On a scenario assessment tool for the operation of the JT-60SA superconducting tokamak

JT-60SA is a superconducting tokamak to be assembled and operated at the JAEA laboratories in Naka (Japan). The tokamak is designed, manufactured and operated under the funding of the Broader Approach Agreement (between the government of Japan and the European Commission) and of the Japan Fusion National Programme; JT-60SA aims to prepare, support and complement the ITER experimental programme. The European contribution to the JT-60SA is, for a large fraction, procured by France, Germany, Italy, Spain and Belgium.This paper summarizes the activities carried out at F4E to develop a user-friendly software tool able to assess in real-time if an operational scenario could be structurally withstood by the magnet system of JT-60SA. Such tool is based on a theoretical formulation which is supported by a series of dedicated finite element method (FEM) calculations, and is able to provide a comparative assessment of any candidate scenario with respect to the baseline scenarios, and a quantitative assessment of all electro magnetic (EM) forces acting on the magnet system at any time during the candidate scenario. The tool as it is presented is specifically designed to be used for the JT-60SA tokamak, though it is designed so to that its usage could be extended easily to any other tokamak.     

Quench propagation and quench detection in the TF system of JT-60SA

In the framework of the JT-60SA project, France and Italy will provide to JAEA 18 Toroidal Field (TF) coils including NbTi cable-in-conduit conductors. During the tokamak operation, these coils could experience a quench, an incidental event corresponding to the irreversible transition from superconducting state to normal resistive state. Starting from a localized disturbance, the normal zone propagates along the conductor and dissipates a large energy due to Joule heating, which can cause irreversible damages.The detection has to be fast enough (a few seconds) to trigger the current discharge, so as to dump the stored magnetic energy into an external resistor. The JT-60SA primary quench detection system will be based on voltage measurements, which are the most rapid technology. The features of the detection system must be adjusted so as to detect the most probable quenches, while avoiding inopportune fast safety discharges. This requires a reliable simulation of the early quench propagation, performed in this study with the Gandalf code.The conductor temperature reached during the current discharge must be kept under a maximal value, according to the hot spot criterion. In the present study, a hot spot criterion temperature of 150 K was taken into account and the role of each conductor component (strands, helium and conduit) was analyzed. The detection parameters were then investigated for different hypotheses regarding the quench initiation.     

Parametric thermo-hydraulic analysis of the TF system of JT-60SA during fast discharge

The evolution of the conductor temperature and of the helium pressure of the central pancake of the TF superconducting magnet of the JT-60SA tokamak in a quench scenario are here discussed. The quench is triggered by a heat disturbance applied at the end of burning and followed by a fast safety discharge. A parametric study aimed at assessing the robustness of the calculation is also addressed with special regard to the voltage threshold, used to define the occurrence of the quench, and to the time delay, that cover all the possible delays in the fast discharge after quench detection. Finally, due to sensitivity analyses the influences of different parameters were assessed: the material properties of the strands (RRR, copper fraction), the magnitude and the spatial length of the triggering disturbance and the magnetic field distribution. The numerical evaluations were performed in the framework of the Broader Approach Agreement in collaboration with CEA, ENEA and the JT-60SA European Home Team using the 1D code Gandalf [1].     

Studies on the requirements of the power supply system for the resistive-wall-mode control in JT-60SA

The JT-60SA satellite tokamak will be built in Naka, Japan. One of the main aims of this machine is to achieve steady-state high-beta plasmas. To reach this result, passive stabilizing plate (SP) and resistive wall modes (RWM) active control system based on 18 in-vessel coils will be installed. In the present design, these coils are placed on the plasma side of the SP, behind the first wall. This solution maximizes the efficiency in producing fast magnetic fields into the plasma by minimizing the shielding effect of the passive structures. Then, if the power supply (PS) and the control system have sufficient dynamic performance, it is possible to control the RWM with very low magnetic fields. This allows minimizing the Ampere-turns and the power requested to control the RWM. Conversely, the very fast dynamics required represents one of the main issues for the design of the RWM control system. This paper, after having recalled the main specification data for the RWM control system deriving from the physics studies, describes the analyses performed to complete the set of requirements necessary for the PS design. The characterization of coils and feeders is shown and the voltage necessary to produce the required current and bandwidth is quantified. Possible connections among PS and coils are analyzed in order to achieve the highest possible flexibility in controlling the RWM with a reduced set of independent PS. Finally, considerations on reasonable voltage margins to cope with load uncertainties are given.     

Suppression of Neoclassical Tearing Modes Towards Stationary High-Beta Plasmas in JT-60U

Results from stabilization of neoclassical tearing modes （NTMs） in JT-60U are described. NTM stabilization and confinement improvement have been demonstrated by employing a real-time NTM stabilization system, where the identification of the location of an NTM and the optimization of the injection angle of the electron cyclotron wave are performed in real time. Also, a high-beta plasma with the normalized beta of 3 has been sustained by suppressing NTM by applying the electron cyclotron current drive （ECCD） before the onset （＇preemptive ECCD＇）. In addition, a simulation code for analysis of the NTM evolution has been developed by combining the modified Rutherford equation with the transport code TOPICS. It is found that the simulation well reproduces the NTM behavior in JT-60U. The simulation also shows that the ECCD width is also important for NTM stabilization, and that the EC wave power for complete stabilization can be reduced by narrowing the ECCD profile.     

Design and Related Experiments of JT-60 Divertor Coils

The JT-60 divertor coils produce a separatrix configuration in divertor operations of JT-60. A suitable separatrix configuration was obtained for a plasma current of 2.1 MA with coil ampere turns of ± 0.755 MAT. A high primary membrane stress of 52 MPa was permissible at the welded joints of the copper conductor made on the site. The mechanical strength of the joints welded in a factory was also improved by means of a press treatment. Electric insulation materials were selected considering a degradation of with stand voltage characteristics due to high cyclic mechanical strain. Vacuum-tight coil cases were composed of rigid rings and U-shaped bellows made of Inconel-625 alloy, and designed to withstand plasma disruption with a current decay time constant of 3 ms. The maximum temperature of the conductor in the periodic operation of divertor discharges was below 155°C which was the allowable temperature of the coil insulation. Molybdenum armor plates coated with titanium carbide and Inconel-625 bellows cover plates were attached against high heat flux from plasma. Thermal and mechanical load tests were carried out using component models to evaluate their performance in advance of the final fabrication of the actual coils. The satisfactory performance of the divertor coils were demonstrated in the pre-operational power test.