Superconducting Toroidal Field Coils Research Articles

Shielding concept and neutronic assessment of the DEMO lower remote handling and pumping ports

Abstract Within the EUROfusion Power Plant Physics and Technology Department the DEMOnstrational fusion power plant (DEMO) is being developed. One of the fundamental challenges is the integration of ports in the vacuum vessel. The lower port of the DEMO machine is particularly challenging due to tight space constraints imposed by the toroidal field (TF) coils and the requirement to provide a large open duct through both the divertor and inside the port to enable for vacuum pumping. In addition, feeding pipes of divertor and tritium breeding blanket need to be integrated and access space must be provided for various remote handling operations. Several neutronics requirements need to be fulfilled, e.g. the nuclear heating of the superconducting TF coils and the gamma radiation levels inside the cryostat need to be limited to reduce occupational exposure to personnel during maintenance, and the irradiation damage and neutron heating in different components need to be considered in the design and limited. The results of neutronic analyses show that further shielding optimization is needed as maximum TF coil heating is still 5× the design limit and the SDDR values orders of magnitude above the target values inside the lower port duct. With this in mind the direction of future design developments is discussed.

Completion of ENEA’s casing procurement forJT-60SA toroidal field coils

In the framework of the Broader Approach program, ENEA is in charge of supplying the Toroidal Field (TF) coil casings for JT-60SA tokamak.ENEA commissioned the manufacture of the full set of eighteen casings for the integration of the TF coils plus two additional spare casings to the company Walter Tosto (WT, Chieti, Italy). The main casing components are one outboard straight leg, an outboard curved leg and three inboard covers. The preliminary design of the casing components has been coordinated by Fusion for Energy (F4E) and the following detail design has been finalized involving the industrial companies in charge of the subsequent integration of the coils.The principal milestones of the five year contract are here summarized. The contract started in 2012 with the realization of two sets of mock-ups representative of the principal casings cross sections of the two legs. In 2013 detail design of the casing components was completed; qualifications of special manufacturing processes were performed and manufacturing procedures were defined. Casing manufacturing activities started in 2014. Seven casings were manufactured during 2015, other seven casings in 2016 and the last four casings were delivered in the first quarter of 2017. The full procurement was then completed with the fabrication of two additional casings in July 2017.This paper provides an overview of the casing procurement describing the production process and manufacturing improvements, reporting the satisfactory acceptance results confirmed by the subsequent proper integration of the superconducting TF coils.

Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR)≥1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

Stress Analyses of the Support Structure and Winding Pack of the Superconducting TF Coil in National Centralized Tokamak

The design activity of the support structure of Toroidal Field (TF) coil is progressed for the National Centralized Tokamak (NCT). For this device, the new support structures of the TF coil is proposed. In the new design, connection keys between coil case and shear panel are adopted in order to decrease the movement of shear panel from coil case which induces the displacement of TF coil during coil operation. For the stress analysis of new structure, FEM model with full-solid elements are developed to evaluate the behavior around the connection keys. The electromagnetic force induces strain and stress on the winding pack which affect the performance of superconductor such as Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Al that is adopted for TF coil in NCT. Therefore, longitudinal strain and transverse stress on the winding pack during coil operation are inspected. For the new design of TF coil, the maximum displacement is 6.4 mm around the upper shear panel and maximum Tresca stress with 415 MPa appears on the shear panel at the midplane. This value is within the allowable stress of SS316LN at 4 K, 547 MPa. For the winding pack, longitudinal strain is evaluated with -0.03 to +0.07%, and transverse stress is evaluated with <40 MPa at the place with the maximum field. These values are too small to affect the critical current of Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3 </sub> Al conductor. Therefore, it is confirmed that the stable operation of superconductor is insured on the new structure of TF coil

Neutronics design of the low aspect ratio tokamak reactor, VECTOR

A neutronic study has been carried out for VECTOR, which is a low aspect ratio tokamak with the aspect ratio of 2.3, to achieve simultaneously sufficient shielding for super-conducting toroidal field coils (TFCs) and tritium breeding ratio (TBR) larger than unity, using the three-dimensional Monte Carlo neutron transport code MCNP-4C. The tritium production is mainly made by outboard blankets consisting of the self-cooled lithium lead (LiPb) as tritium breeding and neutron multiplying material, and the reduced activation ferritic steel F82H as structural material. The inboard blanket consisting of vanadium hydrate (VH 2) and F82H as structural material plays the role of radiation shield for the TFCs. Adding Be layers to the LiPb layers effectively improves the TBR. By employing a 0.8 m thick inboard blanket with a 0.13 m thick LiPb layer, sufficient shielding for the inboard TFC and a total TBR of 1.08 are achieved.

Evaluation of nuclear heating, tritium breeding and shielding efficiency of the DEMO HCLL breeder blanket

Three-dimensional radiation transport calculations employing the MCNP Monte Carlo code have been performed to predict the neutronics performance of the Helium Cooled Lithium Lead (HCLL) DEMO-type blanket. The nuclear energy deposition as well as its spatial distribution, both in radial and poloidal directions, in each blanket module have been calculated. The tritium breeding ratio has been assessed considering different 6Li enrichments and radial depths of the breeder zone of the blanket and the cases providing tritium self-sufficiency have been identified. The shielding performance of the blanket has been investigated at the inboard torus mid-plane regarding the radiation load to the super-conducting toroidal field coil. Appropriate shield configurations are proposed to ensure the radiation load is below the design limit.

Operating Characteristics of the KSTAR Superconducting TF Coil

The Korea Superconducting Tokamak Advanced Research (KSTAR) device, a steady-state-capable advanced Tokamak, is to be built in Korea. The KSTAR device is made up of 16 toroidal field (TF) and 14 poloidal field (PF) superconducting coils. A simulation code for the transient operation of the KSTAR TF and PF coils is developed. For stable Tokamak operation, the operating characteristics of the KSTAR superconducting magnet system are studied for operation scenario. We calculated the temperature margin of superconducting cables, cryogenic flow parameters and the heat deposition in TF and PF coils. The effect of the 3D heat conduction and the eddy current loss in the supporting structure and vacuum chamber of KSTAR is also included.

Effect of plasma disruption on superconducting magnet system in TRIAM-1M

The high field superconducting toroidal field (TF) coils in TRIAM-1M have been successfully operated for more than 10 years. The effect of the plasma disruption on the superconducting TF coils has been studied, since the plasma disruption can give a serious effect such as coil quench. In the actual plasma disruption, the alternative current (AC) loss of the superconducting TF coils increased in proportion to the square of the plasma current just before the disruption. The dependence of the AC loss on the plasma current is in good agreement with the calculation. The AC loss of the TF coil was also measured using external pulse coils with three kinds of characteristic times of variation in the coil current (0.06, 0.5 and 6 ms). The AC loss increased in proportion to the square of the total current of the external pulse coil. The increase in the AC loss depends on the characteristic time of the magnetic field variation. It means that the AC loss due to the disruption can be reduced using the dependence of the AC loss on the characteristic time of the magnetic field variation.

The Spherical Tokamak Path to Fusion Power

The low-aspect-ratio tokamak or spherical torus (ST) approach offers the two key elements needed to enable magnetic confinement fusion to make the transition from a government-funded research program to the commercial marketplace: a low-cost, low-power, small-size market entry vehicle and a strong economy of scale in larger devices. Within the ST concept, a very small device (A = 1.4, major radius ~1 m, similar size to the DIII-D tokamak) could be built that would produce ~800 MW(thermal), 200 MW(net electric) and would have a gain, defined as QPLANT = (gross electric power/recirculating power), of ~2. Such a device would have all the operating systems and features of a power plant and would therefore be acceptable as a pilot plant, even though the cost of electricity would not be competitive. The ratio of fusion power to copper toroidal field (TF) coil dissipation rises quickly with device size (like R3 to R4, depending on what is held constant) and can lead to 4-GW(thermal) power plants with QPLANT = 4 to 5 but which remain a factor of 3 smaller than superconducting tokamak power plants. Large ST power plants might be able to burn the advanced fuel D-He3 if the copper TF coil is replaced by a superconducting TF coil and suitable shield. These elements of a commercialization strategy are of particular importance to the U.S. fusion program in which any initial nongovernment financial participation demands a low-cost entry vehicle.The ability to pursue this line of fusion development requires certain advances and demonstrations that are probable. Stability calculations support a specific advantage of low aspect ratio in high beta that would allow simultaneously βT ~ 60% and 90% bootstrap current fraction (Ip ~ 15 MA, κ = 3). Steady-state current drive requirements are then manageable. The high beta capability means the fusion power density can be so high that neutron wall loading at the blanket, rather than plasma physics, becomes the critical design restriction.

Study of the Tokamak-15 Superconducting Toroidal Field Coil (STFC) Heating Under the Quench

Experiments in Tokamak-15 were performed to study the STFC heating under the quench. The quench was specially caused by current introduction into STFC at the unchanged input helium temperature. The experimental results and simulation data on temperature heating and amount of heat realized in the pancakes under the quench are given. In the experiments was shown that quench occurs in the internal turns of pancakes and estimations of maximal temperature heating corresponds to calculated ones.

Design and installation of instruments in the Westinghouse LCP coil

The Westinghouse Electric Corporation completed the Nb 3 Sn Superconducting toroidal field coil for the Large Coil Program (LCP) which is being managed by the Oak Ridge National Laboratory (ORNL). The coil design which features a force flow cooled conductor and a distributed aluminum plate structure has been described elsewhere [1,2,3,4]. A large number of instruments were installed in the coil because it is a first of its kind proof of concept device. This paper describes the installation of various sensors in the Westinghouse LCP coil.

Large coil task of IEA in the development of superconducting magnets for fusion

The huge superconducting magnets will be used in the toroidal field (TF) coils of the next generation Tokamaks like the INTOR. Therefore, the development of the large superconducting TF coils is urgently needed around the world. Under the auspices of the International Energy Agency, IEA, the big project about that now is going on and is called "The Large Coil Task, LCT" The project is carried out within an IEA-Implementing Agreement signed by EURATOM USA, Switzerland and Japan. US-DOE will act as an operating agent and will produce three coils. A facility for testing six coils in a toroidal array is constructed at ORNL. EURATOM, Switzerland and Japan will produce one coil each for testing with US coils. The test coils are designed to be a 2.5 × 3.5 m bore in D-shaped winding and generate a peak field of 8T. A pair of solenoids will be mounted in the bore of the selected coil to generate a pulsed field for the simulation of the effect of the poloidal field on the TF coils. Besides the main parameters of the test coil, there are significant differences among the six coil designs: superconductors (NbTi or Nb <inf xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</inf> Sn), structural materials, and cooling modes (pool boiling helium or forced flow cooling with supercritical helium gas). The test facility will provide the test stand, vacuum vessel, cryogenics, power supply, and data acquisition systems. The first testing with three pool boiling coils is scheduled to start from July 1982 and the full testing with the six coils from a late half of 1983.

Superconducting magnets for toroidal fusion reactors

Fusion reactors will soon be employing superconducting magnets to confine plasma in which deuterium and tritium (D-T) are fused to produce usable energy. At present there is one small confinement experiment with superconducting toroidal field (TF) coils: Tokamak 7 (T-7), in the USSR, which operates at 4 T. By 1983, six different 2.5 × 3.5-m D-shaped coils from six manufacturers in four countries will be assembled in a toroidal array in the Large Coil Test Facility (LCTF) at Oak Ridge National Laboratory (ORNL) for testing at fields up to 8 T. Soon afterwards ELMO Bumpy Torus (EBT-P) will begin operation at Oak Ridge with superconducting TF coils. At the same time there will be tokamaks with superconducting TF coils 2 to 3 m in diameter in the USSR and France. Toroidal field strength in these machines will range from 6-9 T. NbTi and Nb <inf xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</inf> Sn, bath cooling and forced flow, cryostable and metastable - various designs are being tried in this period when this new application of superconductivity is growing and maturing.