Tokamak Toroidal Field Research Articles

Retraction Note to: Tokamak Coils Materials and Toroidal Field Ripples Calculation Using the Comsol Multiphysics

Retraction Note to: Tokamak Coils Materials and Toroidal Field Ripples Calculation Using the Comsol Multiphysics

Analysis of a Possibility to Use Parallel Non-Twisted Stacks of HTS Tapes as Cable in High Current Conductor of Tokamak Toroidal Field Coils

Development of high current cable-in-conduit conductors (CICC) based on HTS tapes aimed for fusion magnets has started recently. The majority of the suggested CICC use as a HTS cable either, twisted stacks of HTS tapes (stacked technology), or some layers of tapes wrapped around a cylindrical core (CORC technology). An idea to use parallel non-twisted stacks of HTS tapes in a cable design is received skeptically by the fusion community, although such a cable design for tokamak toroidal field (TF) coils has some advantages over twisted ones. The twofold benefits is determined by parallel (deviation- θ <; 5°) orientation of toroidal field vector to surface of HTS tape and mutual orientation of background field and operating current vectors along conductor. Due to such orientation the critical current of tape-Ic(B;T) is considerably (in 5-6 times) higher, due to an anisotropy of HTS tape electro-physical properties. In addition a compressive force is oriented in the most beneficial direction. The main arguments against the non-twisted stacks in the cable design is a concern that the favorable condition described above are not the same in every location of TF coil, or that they will not be favorable in all modes of tokamak operation and the time of TF magnet charging is unreasonably long because of the screening current. The analysis carried out shows that the negative effect of screening currents decreases significantly due to a wavelike magnetic field distribution along conductor and that suggested design of HTS CICC is expected to be successful at any tokamak operation scenario.

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.

Numerical Modeling of the Quench Propagation Phase in the JT-60SA TF Coils

In the framework of the European–Japanese project JT-60SA, the quench tests are performed for each of the 18 NbTi superconducting tokamak toroidal field coils (TFC) in the Cold Test Facility at the CEA Saclay, Gif-sur-Yvette, France. While launching these experimental quench tests, the conductor current sharing temperature ( $T_{cs}$ ) is reached by progressively increasing the inlet helium temperature. The quite complex quench dynamics are then observed due to several coupled physical phenomena influencing the quench propagation. In order to better understand the experimental analyses on coils quench behavior, a numerical model has been used, combining the computation code thermal hydraulic and electric analysis of superconducting cables for the longitudinal quench transient modeling along the cable-in-conduit conductor, and an additional interturn thermal coupling model for transverse heat flux modeling. In this paper, several parametric studies will be done, thanks to simplified numerical simulations in order to identify the predominant physical phenomena driving the JT-60SA TFC quench propagation. The analysis focuses on the linear power impact on the quench initiation, the quench propagation dynamics, and the reverse flow effect.

Performance and Analysis of No-Insulation HTS Toroidal Magnet

As Tokamak toroidal field (TF) coils operate at a high magnetic field and experience a variety of transient heating loads, the applicability of no-insulation (NI) techniques to wind the TF coil is taken into consideration here in an effort to secure high current density and high thermal stability. This paper presents the construction and test results of a NI high temperature superconducting (HTS) toroidal magnet. We fabricated the magnet using BSCCO wire manufactured by Sumitomo. The magnet consists of ten racetrack double pancake (DP) coils. Each DP coil was tested separately in liquid nitrogen at 77 K to obtain its critical current and characteristic resistance, then, after assembly and soldering, the magnet was tested at 77 K to measure the variations in voltage and magnetic field during charging, discharging, and current disturbancing. Finally, we discuss the thermal runaway behavior of the toroidal magnet during overcurrent operation as well as the influence of the quenched coil on the other coils in the system.

Joint testing of the 3 Tesla ST40 spherical tokamak toroidal field coil test assembly

Spherical Tokamaks used in magnetic fusion have a small centre-stack by design which causes a higher field on the conductor. ST40 [1] is a 3 Tesla spherical tokamak with a major radius of R=40cm and minor radius of a=26cm being built by Tokamak Energy Ltd. The high toroidal field (TF) requirement requires a wire current of 250kA flowing in each of the 24 limbs totalling 6MA in the centre-stack. Joint testing was used to investigate the interface resistance between the centre-stack and return limbs under a variety of contact pressures, using different shims, different coatings on the conductor and soldering. Initially a DC 400A current source was used and later an air-core pulsed transformer delivering up to 16kA of current was specifically designed and built for these measurements. Results from this paper can be used to predict the behaviour of critical joints in the region at the ends of the centre-stack where current is passed to the TF return limbs. A comparison with TF joint data from the National Spherical Torus Experiment (NSTX) is also given. This will aid in deciding which jointing method to use for ST40 and the required pressure to pre-load the interface.

Mechanical pre-dimensioning and pre-optimization of the tokamaks’ toroidal coils featuring the winding pack layout

The structural integrity of superconducting magnets that are key elements of a fusion reactor must be ensured. At an early design stage relatively simple calculation tools can greatly facilitate design optimization. The main objective of this paper is the mechanical pre-dimensioning of the tokamak toroidal field coils by simple means prior to the global 3D numerical modeling. A semi-analytical calculation tool that reasonably estimates the static strength of the toroidal field coil under the electromagnetic forces at the critical location (inner leg equatorial plane) is described. The novelty of the approach is that it treats not only the massive coil casing but also the winding pack conductor jacket under an essentially 3D stress state. The calculation tool features pre-optimization of the coil winding for graded layered winding layouts. The minimum space (radial built) required for the coil inboard portion that is a key design parameter is defined after possible winding pre-optimization. The procedure has been successfully benchmarked against numerical solutions and has been used for pre-dimensioning the toroidal coils in the frame of the current 2015 DEMO activity.

Thermal analysis of the cryostat feed through for the ITER Tokamak TF feeder

In Tokamaks, the toroidal field (TF) coil feeder is an important component that is used to supply the cryogens and electrical power for the TF coils. As a part of the TF feeder, the cryostat-feed through (CFT) is subject to low temperatures of 9 and 80 K inside and room temperature of 300 K outside. Based on the features of the International Thermonuclear Experimental Reactor TF feeder, the thermal performance of the CFT under the nominal conditions is studied. Taking into account the conductive, convective and radiation heat transfer, the finite element model of the CFT is built. Transient thermal analysis is performed to determine the temperatures of the CFT on the 9th day of cooldown. The model is assessed by comparing the cooling curves of the CFT after 9 days. If the simulation and experimental results are the same, the finite element model can be considered as calibrated. The model predicts that the cooling time will be approximately 26 days and the temperature distribution and heat load of the main components are obtained when the CFT reaches thermal equilibrium. This study provides a valid quantitative characterization of the CFT design.

Commissioning of the Cold Test Facility for the JT-60SA Tokamak Toroidal Field Coils

Commissioning of the Cold Test Facility for the JT-60SA Tokamak Toroidal Field Coils

RETRACTED ARTICLE: Tokamak Coils Materials and Toroidal Field Ripples Calculation Using the Comsol Multiphysics

The ITER superconducting magnet systems consists of four main sub-systems: toroidal field (TF) coils, central solenoid coils, poloidal field coils, and correction coils. Like many other ITER systems, the magnet components are supplied in-kind by six domestic agencies. The technical specifications, manufacturing processes and procedures required to fabricate these components are particularly challenging. The management structure and organization to realize this procurement within the tight ITER construction schedule is very complex. On the other hand, toroidal magnetic field ripple in tokamak is an important issue in plasma equilibrium and stability studies. Toroidal magnetic field is created by toroidal torus with finite number of coils, therefore the field has a ripple in torus space. In this paper, we have reviewed the ITER magnetic coils materials, and also we have estimated the amplitude of TF ripples and its dependence to numbers of coils using the “Comsol Multiphysics” software. The calculations which performed for three: 8, 16 and 32 toroidal coils, indicates that increasing the number of toroidal coils lead to reduction of magnetic field ripple and lead to more stable plasma, but diagnostic access to plasma is reduces.

Status of the cold test facility for the JT-60SA tokamak toroidal field coils

JT-60SA is a fusion experiment which is jointly constructed by Japan and Europe and which shall contribute to the early realization of fusion energy, by providing support to the operation of ITER, and by addressing key physics issues for ITER and DEMO. In order to achieve these goals, the existing JT-60U experiment will be upgraded to JT-60SA by using superconducting coils. The 18 TF coils of the JT-60SA device will be provided by European industry and tested in a Cold Test Facility (CTF) at CEA Saclay. The coils will be tested at the nominal current of 25.7kA and will be cooled with supercritical helium between 5K and 7.5K to check the temperature margin against a quench. The main objective of these tests is to check the TF coils performance and hence mitigate the fabrication risks. The most important components of the facility are: a 11.5m×6.5m large cryostat in which the TF coils will be thermally insulated by vacuum; a 500W helium refrigerator and a valve box to cool the coils down to 5K and circulate 24g/s of supercritical helium through the winding pack and through the casing; a power supply and HTS current leads to energize the coil; the control and instrumentation equipment (sensors, PLC's, supervision system, fast data acquisition system, etc.) and the Magnet Safety System (MSS) that protects the coils in case of quench. The paper will give an overview of the design of this large facility and the status of its realization.

Structural analysis and optimization of the break-out & feeder of the ITER lower vertical stabilization coil

Break-out & feeder is an essential component connecting the ITER lower vertical stabilization (VS) coil to the outside power source. It plays the role of interface between the in-vessel and out-vessel devices and has large influence on the coil performance. Due to the special location of the ITER lower VS coil, the break-out & feeder has to endure severe in-vessel environment such as high temperature, strong magnetic field and neutron irradiation. High temperature and restricted cooling paths can easily make the break-out under high thermal stress. While square crossing with the Tokamak toroidal magnetic field will result in large Lorentz forces in the feeders. Structural analysis of the break-out & feeder shows overlarge thermal stress concentrating in the coil spine where the conductors lead out. The primary stresses in the feeders are also extremely high. Moreover, there is difference in loading in comparison with the coil main body, which is designed to be mainly in compression and with relaxed crack growth issues, the break-out & feeder is not so compressive and will endure large tensile stress in work. Therefore, relieving the high stresses is important for its design. According to the results of further analysis, structure optimizations such as using block structures in the break-out and modifying the feeder supports are proved with good effect.

An initial study of demountable high-temperature superconducting toroidal field magnets for the Vulcan tokamak conceptual design

Recent developments have made it possible to consider high-temperature superconductor (HTS) for the design of tokamak toroidal field (TF) magnet systems, potentially influencing the overall design and maintenance scheme of magnetic fusion energy devices. Initial assessments of the engineering challenges and cryogenic-dependent cost and parameters of a demountable, HTS TF magnet system have been carried out using the Vulcan tokamak conceptual design (R=1.2m, a=0.3m, B0=7T) as a baseline. Jointed at the midplane to allow vertical removal of the primary vacuum vessel and routine maintenance of core components, structural D-shaped steel support cases provide cryogenic cooling for internally routed YBCO superconducting cables. The cables are constructed by layering ∼50μm thick commercially available YBCO tape, and the interlocking steel support cases self align during assembly to form internal resistive joints between YBCO cables. It is found that designing the TF magnet system for operation between 10K and 20K minimizes the total capital and operating cost. Since YBCO is radiation-sensitive, Monte Carlo simulation is used to study advanced shielding materials compatible with the small size of Vulcan. An adequate shield is determined to be 10cm of zirconium borohydride, which reduces the nuclear heating of the TF coils by a factor of 11.5 and increases the YBCO tape lifetime from two calendar years in the unshielded case to 42 calendar years in the shielded case. Although this initial study presents a plausible conceptual design, future engineering work will be required to develop realistic design solutions for the TF joints, support structure, and cryogenic system.

Mechanical Tests of the ITER Toroidal Field Model Coil

The International Thermonuclear Experimental Reactor (ITER) toroidal field model coil (TFMC) was designed to allow an overall mechanical test representative of the ITER toroidal field (TF) coils design principles and features. The coil, manufactured by the European industry, was tested in two phases at the TOSKA facility of the Forschungszentrum Karlsruhe up to its maximum current of 80 kA. In the single coil tests performed in 2001 the coil was submitted to in-plane loading only, whereas in the two coil tests performed in 2002 the coil experienced also out-of-plane loading, representative of the coil load conditions in the ITER tokamak TF magnet. The paper details the mechanical tests performed, compares them to model predictions and discusses the experience gained in the mechanical behavior of the ITER TF coils.

Properties of impregnation resin added to multifunctional epoxy at cryogenic temperature

AbstractA multifunctional epoxy tetraglycidyl dibenzmethyldiamine and triglycidyl benzylamine were blended, respectively, into impregnation resin, which mainly consists of diglycidyl ether of bisphenol A (DGEBA) and acid anhydride hardener. This resin is expected to impregnate HT‐7U superconducting Tokamak toroidal field coils and improve the fracture toughness of DGEBA–acid anhydride resin at cryogenic temperature. However, the experimental results reveal that resin lap shear strengths decrease remarkably both at ambient and liquid nitrogen temperatures. After blending multifunctional epoxy for several hours, its viscosity increased quickly at room temperature. The usable potting life is too short to impregnate large coils such as those used in fusion reaction. By FTIR the possible reason was investigated. © 2003 Wiley Periodicals, Inc. J Appl Polym Sci 89: 1385–1389, 2003

Overview of Australian activities of fusion neutronics

The new status of the H-1NF Heliac Stellarator as a National Facility and the signed international Implementing Agreement on Collaboration in the Development of the Stellarator Concept should together be a significant encouragement for further fusion research in Australia. In this report the future of fusion research in Australia is discussed with special attention being paid to the importance of Stellarator Power Plant Studies and in particular stellarator fusion neutronics. The main differences between tokamak and stellarator neutronics analyses are identified, namely the neutron wall loading, geometrical modelling and total heating in in-vessel reactor components including toroidal field (TF) coils. An approach to stellarator (TF) coils heating calculations is discussed. This approach is a modification of a previously reported method of total heating calculations in tokamak TF coils. Due to the more complicated nature of stellarator neutronics analyses, simplified approaches to fusion neutronics already developed for tokamaks are expected to be even more important and widely used for designing a Conceptual Stellarator Power Plant.

Effects of CT injector acceleration electrode configuration on tokamak penetration

Through compact toroid (CT) injection experiments on the TEXT-U tokamak (with BT ≃ 10 kG and IP ≃ 100 kA), it has been shown that theacceleration electrode configuration, particularly in the vicinity of the toroidal field (TF) coils of the tokamak, has a strong effect on penetrationperformance. In initial experiments, premature stopping of CTs within theinjector was seen at anomalously low TF strengths. Two modifications werefound to greatly improve performance: (a) removal of a section of the innerelectrode and (b) increased diameter of the `drift tube' (which guides theCT into the tokamak after acceleration). It is proposed that the primary drag mechanism slowing CTs is toroidal flux trapping, which occurs when a CT displaces transverse TF trapped within the flux conserving walls of the acceleration electrodes (ordrift tube). For a simple two dimensional (2-D) geometry, a magnetostatic analysis produces a CT kinetic energy requirement of 1/2ρv2 ⩾ α(B02/2μ0), with α = 2/(1-a2/R2) adimensionless number that is dependent on the CT radius a normalized by the drift tube radius R. For a typical CT, this can greatly increase the required energies. A numerical analysis in 3-D confirms the analytical result for long CTs (with length L such that L/a ≳ 10). In addition to flux trapping, the CT shape is also shown to affect theenergy criterion. These findings indicate that a realistic assessment of the kinetic energyrequired for a CT to penetrate a particular tokamak TF must take intoaccount the interaction of the magnetic field with the electrode walls of the injector.

Diamagnetic Measurement of JFT–2 Plasma Heated by Neutral Beam Injection

A neutral beam was injected into the plasma in the JFT–2 tokamak, and the poloidal beta value βp of the plasma was determined by a diamagnetic method in which the change in the magnetic flux due to the plasma was obtained by measuring the very small perturbation of the current in the tokamak's toroidal field coil. The ratio of the perturbed to unperturbed currents in the coil was found to be (2–3)×10-4. The poloidal beta value βpd determined by this method agrees within experimental error with that obtained from magnetic and energy profile analyses. βpd increases linearly with the total power Pnet deposited by the neutral beam in the plasma when Pnet=1.5 MW. The heating efficiency of the beam injection heating was found to be lower than that of Joule heating.

Structural response of Tokamak TF magnet systems to in-plane and out-of-plane electromechanical loads

A finite thickness, hybrid toroidal shell is presented to model the structural behavior of a Tokamak toroidal field magnet system, consisting of a discrete set of toroidal coils, with intercoil support structure. Both in-plane and out-of-plane electromechanical loads are considered. When the stiffness of the intercoil structure is accounted for, the membrane formulation of the model demonstrates strain incompatibility in the funicular “D” coil configuration. Bending effects are predicted in this configuration; this is also demonstrated numerically. The model is applied in parametric structural analysis of several TF coil configurations with the aspect ratio of the 1981 FED design, under the action of in-plane and out-of-plane loads. An elliptical TF coil shape which exhibits significantly smaller bending stresses than the “D” configuration is recommended as an alternative when, to meet plasma engineering requirements, an elongated coil shape is necessary.

Elastic buckling of superconducting Yin-Yang magnets for fusion

A small superconducting model of yin-yang coils for the Mirror Fusion Test Facility has exhibited a magnetoelastic buckling instability. The 1:22 model, wound with 150 turns of niobium titanium wire and potted in epoxy, can carry up to 30 10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sup> Amp-turns. Rotational motion of the magnets is observed to increase as the current approaches a critical value, while the torsional natural frequency is observed to decrease with current. A prediction of the buckling current based on a theoretical model is found to be higher than the measured values. The phenomenon is similar to that observed for tokamak toroidal field magnets. Structural design implications for Yin-Yang magnets are discussed.