Central Solenoid Research Articles

Analysis of the Mechanical Behavior of the KSTAR CS Magnet During Long-Pulse Plasma Discharges

The KSTAR central solenoid (CS) magnet consists of four pairs of coils stacked vertically and eight preloading structures that compress the coils in the axial direction. The strain and dis-placement of the CS magnet are monitored in real time, limiting the minimum compression as an important parameter of the supervisory interlock system for the safe operation of KSTAR. An equivalent vertical force method for a simplified multi-spring system was developed to analyze the preload and displacement variations of the CS magnet. The equivalent vertical force is based on the plasma and poloidal field (PF) coils currents during plasma discharge. The estimated displacements by the developed method agreed fairly well with the measured displacements of the long-pulse plasma discharges in 2017 (#18437) and 2018 (#21735) campaigns. The analyzed displacement can be used to simply estimate the remaining preload of the CS coils. This approach will contribute to the development of advanced operation scenarios for large plasma current and long-pulse discharges. By applying this algorithm to the plasma control system, the CS magnet can be controlled more safely during KSTAR operation.

AC Loss Measurement of the DEMO TF React&Wind Conductor Prototype No. 2

The EUROfusion DEMO is being designed as the fusion machine to be built after ITER. During the preconceptual phase, several design options are investigated by theoretical analyses as well as tests on newly developed conductor prototypes. One design option for the toroidal field magnet (TF) and central solenoid (CS) is based on flat Nb 3 Sn forced-flow conductors made with react&wind technique. The usage of these conductors simplifies some steps in the conductor and coil manufacturing, and together with graded layer-winding allows approx. 50% reduction of the required amount of Nb 3 Sn compared to ITER-like design based on wind-react-insulate pancake-winding. Two full-size prototype cables for DEMO TF coil were manufactured, jacketed and tested in several test campaigns in SULTAN test facility. The DC results for the second prototype, RW2, rated for 63 kA at 12.3 T, were presented and published in 2018. The AC loss data were collected over several test campaigns performed on various assemblies of RW2 at AC field parallel and perpendicular to the broad cable side. The measurements done at 4.5 K and 20 K allow us to decompose the AC loss contributions originating from the bundle of superconducting strands (hysteresis and coupling loss) and copper-matrix stabilizer located around the cable. The AC loss for sinusoidal and trapezoidal field variations are presented and discussed. The low AC loss of the flat cable makes the cable an attractive choice for the central solenoid operating in a pulse mode.

Magnetostructural Calculations and Design Study of the DTT Central Solenoid

The Divertor Tokamak Test facility (DTT) is a project of an experimental tokamak reactor developed in Italy, in the framework of the European Fusion Roadmap. This work presents the magnetic and the structural assessment of the performance of the DTT central solenoid. The CS is the core magnet of the poloidal system and generates the magnetic flux needed to induce the plasma current. This magnet is composed of a stack of six layer-wound independently energized modules, comprised of Nb3Sn Cable-in-Conduit Conductors. To optimize the amount of superconductive material, each module is divided into two submodules. The inner- most submodule operates in a range of about 8/13.5 T, while the outer one at 6/8.5 T. The objective of the design process is to obtain a coil that is capable of providing the required magnetic performance while being structurally compliant. To address this problem, an analytical assessment has been carried out and a thoroughly parametric Finite Element Model (FEM) has been implemented.

Modelling quench of a 50 kA REBCO conductor with soldered-twisted-stacked-tape-cable strands

A 50 kA REBCO conductor based on the Twisted-Stacked-Tape-cable concept is being developed at Swiss Plasma Center (SPC) for EU-DEMO central solenoid. The conductor contains 12 strands and each one is made of twisted REBCO tape stack soldered in round copper profile with rectangular slot. In this paper, first it is verified if the strands could be modelled as monolithic elements for quench simulation, which means instantaneous heat propagation within the domain and homogeneous current density, by studying the influence of the inter-tape and stack-profile resistance on their heat and current transfer. Then a local disturbance induced quench on the 50 kA REBCO conductor is studied in detail considering various inter-strand electrical resistance.

Demountable Coaxial Clamped Joint For ITER Central Solenoid Module Final Test Program

The ITER Central Solenoid Modules (CSM) are being fabricated at General Atomics (GA) at their Magnet Technologies Center in Poway, CA. Each of the seven modules will undergo final testing at the GA facility to demonstrate their performance at 48.5 kA, 11 T, and at 4.5 K. In order to perform tests on multiple modules, a demountable coaxial joint using indium wires was developed to connect the modules to the feeders. Two full scale joints were fabricated to establish the assembly technique and preliminarily tested (up to 3 kA) on the CSM qualification coil (Cu conductor). The same joint components were then tested using a superconducting jumper (NbTi) and their resistance measured (2.3 and 4.9 nΩ, at 4.5 K, 40 kA). In parallel to measuring the voltage across the joints, a calorimetric evaluation was also performed and it confirmed the low resistance measurements of both joints. Testing of the same coaxial clamped joints on a heat treated Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn jumper took place in February 2019 and the results are reported in this paper. With these acceptable results, the same joint components and assembly techniques will be utilized to connect the 14 joints on the seven CSM (Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn) for full current testing.

A Reliability Analysis of CFETR CSMC Heat Treatment System Based on RPN-HAZOP Method

The safe and reliable operation of the China Fusion Engineering Test Reactor (CFETR) Central Solenoid Model Coil (CSMC) heat treatment system is the basis for the establishment of the CFETR central solenoid, Nb3Sn superconducting coil. The purpose of safety analysis of a heat treatment system is to find out the unreliable factors of the system, analyze the weaknesses of the system, predict the possibility and severity of system failure, and then improve the reliability of the system. The method of combining risk priority number (RPN) with Hazard and operability study (HAZOP) can make up for the deficiency of quantitative information in traditional HAZOP analysis. It is concluded that the fan subsystem is the weakest part of the heat treatment system by means of RPN-HAZOP. The deviation caused by temperature, cleanliness, and impurity gas content will greatly affect the performance of the superconducting coil. The author tries to put forward scientific and reasonable measures according to the principle of giving priority to solving the principal contradiction. Guided by the measures and suggestions, the heat treatment system runs normally, safely, and reliably. The results of the experiment show that all the process indexes meet the requirements of coil heat treatment technology.

The Necessary Parameters Calculation for CSMC Mechanical Analysis

The Central Solenoid Model Coil is a hybrid superconducting coil being developed in Institute of Plasma Physics Chinese Academy of Sciences. The R&D of Central Solenoid Model Coil will lay the foundation for the fabrication of China Fusion Engineering Test Reactor Central Solenoid coil. One of the main purposes during the R&D of Central Solenoid Model Coil is to verify its mechanical performance under the complicated load cases. In order to efficiently and accurately carry out the mechanical analysis, some necessary parameters such as the electromagnetic state under assembly error, the thermal contact resistance and the equivalent material property of conductor winding packet should be calculated in advance. The electromagnetic state here mainly refers to the asymmetric magnetic field and electromagnetic force. The calculation of magnetic density is based on elliptical integrals, the results will provide guidance for the assembly optimization of coil modules. And the subsequently obtained electromagnetic force will offer load boundary condition for the coil stability analysis. The thermal contact resistance is calculated under the assumption that the interface asperity approximately obeys the Gaussian distribution. The thermal contact resistance will be used in the thermal analysis. And the key advantage is which makes the prediction of preload force variation more accurate during coil the cooling down process. The equivalent material property is calculated by using generalized Hook's law and finite element method. Based on the equivalent material property, the coupling analysis model of CSMC will be significantly simplified. Moreover, it makes CSMC full model analysis under non-uniform loads become possible.

Control and Diagnostic System for CFETR CSMC Testing Platform

The central solenoid model coil (CSMC) of the China Fusion Engineering Testing Reactor (CFETR) is currently in the manufacturing phase. Once completed, the CSMC will be tested on a testing platform. The testing platform is composed of several subsystems, which are coordinated by a control system to conduct a successful experiment and collect scientific data generated during the experiments. A diagnostic system is employed for the purpose of monitoring the status of a superconducting magnet and triggers the alarm system if an abnormal signal has been detected. The central control system of the CSMC testing platform is called the control data access and communication (CODAC) system. In order to realize the normal data communication between the CODAC system and a subsystem, an experimental physics and industrial control system (EPICS) Channel Access (CA) protocol is needed.

First Results and Plasma Current Start-Up in Taban Tokamak

Plasma current start-up and pre-ionization is an important motivation in Taban tokamak. The plasma current start-up has been successfully achieved using a central solenoid and a pre-ionization system and the experimental conditions have been optimized. Two distinct schemes were used to start-up the plasma current. Hot cathode emitter and electron cyclotron resonance (ECR) was used as the pre-ionization system. ECR pre-ionization circuit of 2.45 GHz magnetron was modified for long duration pulse mode. The transient plasma parameters were measured during discharge as a function of time in different conditions, while a toroidal field of 0.1 T with 4 ms flat-top was being applied. The experimental results showed that the plasma current could be started up by both ECR and hot cathode emitter. In the optimum conditions, a plasma current of 7.5 kA has been achieved with pulse duration of ~ 0.7 ms using the electron cyclotron pre-ionization system. Under such circumstances, the X-ray emission, as well as, the impurity level was minimized.

Start-Up Studies of GLAST-III Spherical Tokamak in the Presence of Poloidal Field

GLAST-III (glass spherical tokamak) is a small device having an aspect ratio of two (R/a = 2) with major radius R = 20 cm and minor radius a = 10 cm. Experiments are performed to examine the effect of increasing the poloidal field (PF) produced by poloidal coils as well as to study the effect of the inductance of PF coils on the plasma current formation. To investigate the GLAST-III plasma, several diagnostic tools such as Rogowski coil, loop voltage, photodiode, Optical emission spectroscopy, high-speed camera, and Langmuir probe are used. After optimizing the working pressure, the first experiment is performed using toroidal field coils and central solenoid along with radio frequency (RF) pre-ionization source. A plasma current of 1 kA is achieved for 1.2 ms. In later experiments, the effect of the poloidal magnetic field is also included. It is observed that with increasing the PF, the plasma current increases, attain a maximum value (5 kA) and then after a critical value of the applied poloidal magnetic field, the plasma current starts decreasing. Similarly, with the inclusion of inductor in series with the PF coils, the plasma current value is reduced and its pulse duration is increased. Both optical emission and Langmuir probe diagnostics results show a similar trend of the plasma current.

Conceptual design of Japan’s fusion DEMO reactor (JADEMO) and superconducting coil issues

Goals of Japan’s fusion demonstration (DEMO) reactor are to demonstrate (1) steady and stable electric power generation in a power plant scale, (2) self-sufficient production of fuel (tritium), and (3) reasonable availability using a remote maintenance scheme anticipated in a commercial plant. Main design parameters of JA DEMO are a plasma major radius of 8.5 m, fusion output of 1.5-2 GW, magnetic field on the plasma axis of 5.94 T. The superconducting coil system of the reactor consists of a central solenoid (CS), 7 poloidal field (PF) coils and 16 toroidal field (TF) coils. Regarding CS and PF coils, superconducting coil technology on DEMO is basically the same as that on the world largest fusion experimental reactor called ITER. In contrast, TF coils have a technology gap on magnetic energy and the resulting stress between ITER and DEMO due to their size and magnetic field. In particular, the necessity of higher design stress is critical for TF coils, requiring the development of high strength cryogenic steels surpassing the existing ones. The fundamental design strategy to mitigate tolerances in TF coil fabrication is also presented.

Design and Analysis of the US ITER Central Solenoid Module Lifting Fixture

The central solenoid (CS) consists of six large high field superconducting magnets (also known as modules) approximately 4 m in diameter and 2 m tall that weigh approximately 120 tonnes each. These large and complex modules create challenges during assembly of the CS that require the development of custom assembly tooling such as the CS lifting fixture. The CS module lifting fixture is designed to lift and stack the six CS modules in the assembly building on the ITER site. Because of its unique design, fabrication, and assembly features, no lifting attachments could be incorporated within or under the CS modules. This limitation motivated the development of a friction-based lift fixture. The design and evaluation of the CS module lifting fixture considered both worker safety and investment protection, and the assessments were performed to international codes and standards. The CS module lifting fixture consists of two principal subassemblies: spider assembly and ring weldment. These subassemblies allow the frictional force to be augmented by the mechanical advantage of shallow-angle wedges. Large radial preloads created by both screw jack assemblies and the weight of a CS module develop frictional forces capable of performing a lift with a safety factor of at least 2. The design effort resulted in the use of low friction linear bearings on angle surfaces to ensure constant pressure, integrated jacks for pretensioning the fixture prior to lifting, and load pin strain gauges for monitoring the normal force. Testing of various materials and surface treatments led to the selection of laminated aluminum and rubber pads as the friction interface on the CS lifting fixture side and a grit-blasted Nitronic 50 stainless steel band on the CS module side. A redundant lifting method using the module slings between the spider and module is also utilized after the initial friction lift. The CS lifting fixture provides a safe and reliable solution for lifting and stacking the CS module during assembly.

DC Test Results of the DEMO TF React&amp;Wind Conductor Prototype No. 2

In 2013, the EUROfusion organization launched the preconceptual design studies of the fusion DEMO tokamak. Several options of superconducting magnet design are investigated within the scope of the project. The magnets based on Nb3Sn, namely the Toroidal Field (TF) and Central Solenoid (CS) coils, can be made either using the Wind&React technology, as in ITER, or React&Wind (RW) technology with potentially significant cost-saving benefits. Within the R&D program of the Swiss Plasma Center, the first full-size DEMO TF RW conductor prototype, RW1, was manufactured and tested in 2015–2016. The second, improved prototype, RW2, was built and tested in 2017–2018. The experimental results collected over several test campaigns performed on four assemblies of RW2 rated for 63 kA at 12.23 T are presented. The current sharing temperature $T_{{{cs}}}$ of the latest RW2 assembly in DEMO operating conditions reached 7.16 K after 1000 electromagnetic cycles and four thermal cycles. This $T_{{{cs}}}$ corresponds to the effective strain ϵeff = −0.27%, significantly exceeding the design requirement of $T_{{{cs}}}\,\geq \,\text{6.7}\,\text{K}$ . If RW2 have been used in ITER TF coils, the $T_{{{cs}}}$ at 68 kA and 10.9 T would be 7.42 K, i.e., approximately 1 K higher than most of the ITER TF samples, and that with only 55% of Nb3Sn used in ITER TF conductor. The tests of RW2 proved that the cost-saving potential of the React&Wind technology is enormous.

Modeling the ITER CS AC Losses Based on the CS Insert Analysis

The cable-in-conduit conductor that will be used for the manufacturing of the ITER central solenoid (CS) modules has undergone a long series of qualification tests: the latest was performed in 2015 at QST, Naka, Japan, on the central solenoid insert (CSI) coil. In this work, the AC losses dataset collected during the CSI test campaign is interpreted using a lumped-parameter model for the coupling and hysteresis losses. The model is first benchmarked against the results of the THELMA code and then, after the implementation in the 4C thermal-hydraulic code, successfully validated against experimental data from tests performed on the CSI. With the validated AC loss model, the predictive analysis of the performance of the ITER CS is then carried out using again the 4C code, both in nominal conditions and with a reduced coolant mass flow rate in the most loaded pancake; it is shown that the minimum temperature margin required by the design is always satisfied, for both virgin (1 K) and cycled (1.5 K) conductor.

Feasibility Analysis of 100 kA DC Commutation Scheme to be Applied in the Quench Protection Unit of CFETR

China fusion engineering test reactor is a new tokamak device, which is one design option under the consideration of the China National Integration Design Group employs superconducting magnets. The main target of this project is to build a self-sufficient fusion engineering tokamak reactor with its fusion power 50-200 MW. Toroidal field coils, center solenoid coils, and poloidal field (PF) coils, whose rated current range from 52.9 to 90 kA, produce the magnetic field for plasma generation and confinement. The core of quench protection unit is to divert the coil current into discharge resistor by using commutation technology, which should be determined by the rated current and voltage requirement of magnets. Different kinds of technologies for quench protection unit (QPU) are considered, and an advanced commutation technology composed of a mechanical bypass switch, vacuum circuit breaker (VCB), and LC commutation circuit based on artificial current zero technology has been worked out to meet the current and voltage requirement (10 kV/100 kA). To provide a faster and more reliable removal of the energy stored in those superconducting coils, this paper analyzes in detail the feasibility of a commutation scheme based on artificial current zero technology by analyzing and simulating each stage of the commutation process at such large rated current and voltage conditions. Both analysis and simulation results verified that the operating reliability of QPU can be guaranteed by this 100 kA dc commutation scheme.

Investigation of fine structure formation of guide field reconnection during merging plasma startup of spherical tokamak in TS-3U

Ion heating/transport and its fine structure formation process through magnetic reconnection have been investigated by high guide field tokamak merging experiments in TS-3 and TS-3U. In addition to the previously reported demonstration of high-temperature plasma startup without center solenoid, the detailed fine structure formation process of reconnection heating has been revealed using new 96CH/320CH ultra-high-resolution 2D ion Doppler tomography diagnostics. By identifying the double-axis field configuration with the X-point on the midplane using in situ magnetic probe diagnostics, the detailed measurement successfully revealed that the ion temperature profile forms two types of characteristic heating structure, both around the X-point and downstream. The former is affected by the Hall effect to form a tilted heating profile, while the latter is affected by the transport process which a forms a poloidal double-ring-like structure. The achieved ion heating mostly depends on the reconnecting component of the magnetic field, and the contribution of the guide field to decrease the heating efficiency tends to be saturated in the high guide field regime. Under the influence of better toroidal confinement with higher guide field, the downstream ion heating is transported vertically, mostly by parallel heat conduction, and finally forms a poloidal ring-like hollow distribution aligned with the closed flux surface at the end of merging.

The differentially-tilted toroidal field coil concept for tokamaks

The implications of the adoption of a tokamak's toroidal field coil characterized by differential tilting in the azimuthal direction are investigated. From an engineering point of view, the major advantage introduced by such coils is a drastic reduction of some components of the electromagnetic forces in certain areas. As a beneficial side-effect, they generate poloidal field, in addition to toroidal field. The former advantage allows for a partial relaxation of the reinforcing structural material required in the machine design, while the poloidal flux generated during the current rise, when used in conjunction with the conventional central solenoid, would allow for discharges of longer duration. This paper presents results obtained by applying the tilting optimization procedure to circular and D-shaped coils, and characterized by geometrical and physical parameters proper of high-field compact tokamaks, since the issue of electromagnetic force reduction is most relevant in these devices.

Calculation method for pulsed magnetic field energy supplied to Nb3Sn ITER CS conductors during SULTAN stability tests

Cable-In-Conduit Conductors (CICCs) for the ITER Central Solenoid (CS) magnets are designed to operate in the presence of fast changing current and magnetic field during the plasma-operating scenario. For ITER, the AC loss of several types of Nb3Sn CICCs was experimentally tested, but only very limited experimental data is available for quantitative analysis of the minimum quench energy (MQE). In the SULTAN testing facility (Swiss Plasma Centre) few CS conductors were tested on MQE, but the magnetic field amplitude and ramp rate settings are far from the actual ITER operating conditions. Nevertheless, such tests are needed as a basis to calibrate and benchmark the codes that describe the quench behavior. Moreover, during the stability tests in Sultan, the temperature measurements show severe fluctuations, which can introduce a large error for the energy calculation. An interpretation is given for the temperature fluctuation and a procedure is proposed to significantly reduce the error in the pulsed energy calculation.

Electromagnetic and thermal stability of the ITER Central Solenoid during a 15 MA plasma scenario

The combination of the electromagnetic conductor model JackPot ACDC with the thermo-hydraulic model THEA is able to reproduce and predict the behavior of cable-in-conduit conductors (CICCs) under any varying current and magnetic field. The combined JackPot-THEA model is used to simulate the turns of the CSU2 quadra-pacake of the ITER Central Solenoid (CS) that are subjected to the most demanding conditions of the 15 MA plasma scenario. The considered section is a pancake with a 150 m length of Nb3Sn CICC from helium inlet to helium outlet, from the inner towards outer turns. The simulation results in terms of temperature and local strand electric field levels are compared to those obtained from earlier minimum quench energy analysis in order to evaluate possible risk of quench. The results confirm sufficient stability of the ITER CS coil, both from electrical and thermal point of view. The temperature evolution and flow of a helium heat slug along the conductor during the 15 MA plasma scenario is also analyzed. It appeared that sequential multiple burn cycles can be carried out without accumulation of temperature in the helium heat slug, essential for continuous energy production for a fusion power plant.

Progress of the JT-60SA project

The JT-60SA project was initiated in June 2007 under the framework of the Broader Approach Agreement and Japanese National Fusion Programme for an early realization of fusion energy by conducting supportive and complementary work for the ITER project towards supporting the basis for DEMO. With the project now in an advanced implementation stage, the early defined approach for its implementation has proven to be successful and hence continues to be employed. This is underpinned by the very close collaboration between QST in Japan, F4E in Europe, and all other European stakeholders: the EU Voluntary Contributors and EUROfusion. As of September 2018, the closure of the torus has been accomplished. All TF coils have been manufactured, tested at full current and cryogenic temperature demonstrating a consistent temperature margin, and assembled. All manufacturing of equilibrium field coils was completed by the middle of August 2016. Three central solenoid (CS) modules were completed by March 2017 while manufacturing of the last CS module has recently been completed. The manufacture and delivery of all 26 high temperature superconductor current leads was completed in November 2017. All large power systems, including the switching network units and super conducting magnet power supplies, have also been manufactured, delivered and with few residual commissioning activities still ongoing in Naka. With the cryostat base already in place since 2013, the cryostat vessel body cylindrical section has been manufactured, preassembled, measured and delivered to Naka. The full scope of the cryoplant, manufacturing and commissioning, is now successfully completed. The final assembly phase has therefore started together with the gradual execution of integrated commissioning leading to the completion of the assembly in March 2020 and a first plasma in September 2020. The efficient start-up and scientific exploitation of JT-60SA by the large international team is a challenging enterprise, which will be similar to, and provide important input to, the ITER start-up phase. To optimize this phase, a broad set of coordinated activities have been carried out over recent years by a joint Japanese-EU JT-60SA research unit, fully integrated in the integrated project team and liaising with the broader Japanese and EU fusion physics community. The paper will provide an overview of the progress of the manufacturing and assembly of the JT-60SA machine towards first plasma, and progress in preparing for the next phases of JT-60SA following this milestone.