Toroidal Field Magnet Research Articles

The first experiments in SST-1

A steady state superconducting tokamak (SST-1) has been commissioned after the successful experimental and engineering validations of its critical sub-systems. During the ‘engineering validation phase’ of SST-1; the cryostat was demonstrated to be leak-tight in all operational scenarios, 80 K thermal shields were demonstrated to be uniformly cooled without regions of ‘thermal runaway and hot spots’, the superconducting toroidal field magnets were demonstrated to be cooled to their nominal operational conditions and charged up to 1.5 T of the field at the major radius. The engineering validations further demonstrated the assembled SST-1 machine shell to be a graded, stress–strain optimized and distributed thermo-mechanical device, apart from the integrated vacuum vessel being validated to be UHV compatible etc. Subsequently, ‘field error components’ in SST-1 were measured to be acceptable towards plasma discharges. A successful breakdown in SST-1 was obtained in SST-1 in June 2013 assisted with electron cyclotron pre-ionization in the second harmonic mode, thus marking the ‘first plasma’ in SST-1 and the arrival of SST-1 into the league of contemporary steady state devices.Subsequent to the first plasma, successful repeatable plasma start-ups with E ∼ 0.4 V m−1, and plasma current in excess of 70 kA for 400 ms assisted with electron cyclotron heating pre-ionization at a field of 1.5 T have so far been achieved in SST-1. Lengthening the plasma pulse duration with lower hybrid current drive, confinement and transport in SST-1 plasmas and magnetohydrodynamic activities typical to large aspect ratio SST-1 discharges are presently being investigated in SST-1. In parallel, SST-1 has uniquely demonstrated reliable cryo-stable high field operation of superconducting TF magnets in the two-phase cooling mode, operation of vapour-cooled current leads with cold gas instead of liquid helium and an order less dc joint resistance in superconducting magnet winding packs with high transport currents. In parallel, SST-1 is also continually getting up-graded with first wall integration, superconducting central solenoid installation and over-loaded MgB2-brass based current leads etc. Phase-1 of SST-1 up-gradation is scheduled by the first half of 2015, after which long pulse plasma experiments in both circular and elongated configurations have been planned in SST-1.

Extended Quality Control Process Applied on TF Strand and TF Conductor Production for JT-60SA Project

In the framework of the JT-60 Super Advanced (JT-60SA) project, as part of its contribution to the Broader Approach Agreement, Europe provides the full toroidal field (TF) magnet system. For this purpose, Fusion for Energy (F4E) is procuring 27 km of TF conductor, which is of cable-in-conduit type, including 486 strands (2/3 NbTi and 1/3 copper) embedded into a rectangular stainless steel jacket. Since the start of procurement in 2010, the totality of strand has been manufactured, and more than half of the total conductor unit lengths have been accepted by F4E. In the course of this material fabrication, a large quality control (QC) program was established to consolidate the robustness of the quality diagnostic of both productions' follow-up. As for the TF strand production, a comprehensive campaign of critical properties evaluation was led to ensure that the T CS (temperature of current sharing) value in the operation condition is compliant with a 1-K temperature margin in the JT-60SA tokamak. From the large amount of data collected over the two-year production period, a statistic analysis is presented in the paper, together with the methodology that was used to build it. As for TF conductor production, the QC database provided hydraulic properties at low Re. In this framework, a covering QC campaign was led to crosscheck the acceptance data and, at the same time, quoting the reliability of acceptance data exploitation for predictive mass flow analyses in operation. An extra QC, including material analyses and quantitative energy-dispersive X-ray (EDX) analysis of both cable and jacket, was also performed. Highlights are mainly put to the scientific approach of the present topics investigated, but together, some considerations on the risk approach are also presented.

Degradation of the Cu Residual Resistivity Ratio in Cr-Plated Composite Nb<sub>3</sub>Sn Wires

Some of the bronze-route Nb3Sn wires produced for the ITER toroidal field magnets have values of the Cu residual resistivity ratio (RRR) lower than the specification (RRR > 100) when reacted with the longest ITER heat treatment cycle (cycle A: high-temperature plateau 200 h at 650 ◦C). As the low RRR value was suspected to be due to the Cr plating, CERN has carried out extensive investigations on two different Cr-plated Nb3Sn wires in order to assess how the RRR is influenced by the Cr plating. Each type of wire was reacted with three heat treatments differing for the duration of the high-temperature plateau (from 100 to 200 h). The presence of Cr in the Cu stabilizer was investigated using a transmission electron microscope together with energy-dispersive X-ray spectrometry. In the most critical wire reacted for 200 h, CrS nanoparticles and Cr in solid solution (0.02 at.%) with Cu were found up to 13 μm from the Cr–Cu interface. At larger distance, we could not quantify the Cr concentrations because of the detector sensitivity limit (about 0.016 at.% Cr in Cu). We provide a model that can explain the measured RRR degradation purely in terms of the Cr contamination that we observed.

Investigation of ITER TF Conductor Hydraulic Resistance

Conductors for winding ITER toroidal field magnets are cable-in-conduit type where multi-stage petals are twisted around central cooling spiral inserted into stainless steel jacket. Every conductor unit length and its components pass various tests during manufacturing and the final acceptance test consists of 760 m or 415 m long conductor global leak test and an investigation of conductor hydrodynamic performances. The hydraulic resistance is one of the main TF CICC's parameter to provide sufficient cool down and stable work of the magnet system during operation. Starting from 2011, when the first copper Dummy conductor was completed, 22 conductors in total were tested in a test facility that was designed and manufactured at National Research Centre “Kurchatov Institute.” A scheme of the pressure drop test facility, measurements technique and results of mass flow rate measurements depending on inlet and outlet pressure difference are presented in this work. Measurements were carried out by flowing nitrogen gas through conductor at room temperature. However, in accordance with the theory of turbulent self-similarity it is possible to develop a simple criterion for estimation of hydraulic resistance at any temperature. The model of two parallel independent flows has been used. The simple explanation of mass flow rate non-linearity at small pressure difference has been provided.

Assessment Studies and Manufacturing Trials for the Conductors of DEMO TF Coils

This paper reports the main outcomes of the 2013 design activities for the toroidal field magnet of the European nuclear fusion reactor DEMOnstration Power Plant (DEMO). Within the current project plan, the manufacture of full-size conductor prototypes during 2014 and their test during 2015 are foreseen, aimed at the assessment of the proposed cable-in-conduit-conductor technologies. Industrial partners have been thus already involved at this early stage, and the preliminary cabling trials performed are illustrated here. As far as the coil design assessment is concerned, one of the main aspects to be investigated is the mechanical performance of the magnet, considering its large size and the high electromagnetic forces involved. An overall 3-D finite-element model of the coil has been set up, and mechanical stress analysis has been carried out for a selected instant of the plasma scenario after the computation of the electromagnetic loads within a full 3-D system model, including central solenoid and poloidal field coil contributions. The resulting loads have been mapped on the finite-element-method model, making use of a novel mesh-independent interpolator based on radial basis functions. The results reported here highlight a quite high level of stress, which is close to the acceptance limits, in some regions of the steel casing. Further investigations and assessments are required, but the most critical areas are clearly identified.

Advanced examination techniques applied to the qualification of critical welds for the ITER correction coils

The ITER correction coils (CCs) consist of three sets of six coils located in between the toroidal (TF) and poloidal field (PF) magnets. The CCs rely on a Cable-in-Conduit Conductor (CICC), whose supercritical cooling at 4.5K is provided by helium inlets and outlets. The assembly of the nozzles to the stainless steel conductor conduit includes fillet welds requiring full penetration through the thickness of the nozzle. Static and cyclic stresses have to be sustained by the inlet welds during operation. The entire volume of helium inlet and outlet welds, that are submitted to the most stringent quality levels of imperfections according to standards in force, is virtually uninspectable with sufficient resolution by conventional or computed radiography or by Ultrasonic Testing. On the other hand, X-ray computed tomography (CT) was successfully applied to inspect the full weld volume of several dozens of helium inlet qualification samples. The extensive use of CT techniques allowed a significant progress in the weld quality of the CC inlets. CT is also a promising technique for inspection of qualification welds of helium inlets of the TF magnets, by far more complex to examine due to their larger dimensions.

R&D on 52-kA HTS Current Lead at ASIPP

The Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) is experienced in the design and fabrication of high-temperature superconductor (HTS) current leads such as those successfully used in the Experimental Advanced Superconducting (EAST) Tokamak. The aim of the 52-kA trial lead was to develop a prototype for the leads supplying the poloidal field and central solenoid magnets of ITER. This lead differs from former HTS current leads and from another trial (68 kA) lead developed in parallel for the toroidal field magnet, in that it features a foil-type heat exchanger, a helium-cooled transition section, and a binary shunt made of stainless steel and phosphor bronze. The 52-kA trial lead was fabricated and tested in the EAST Tokamak facility. Results of the test are presented. An analysis of these results indicates that the design concepts introduced are validated. The test also served to validate the HTS stacks that equipped the lead. These were supplied by InnoST.

Modeling an unmitigated thermal quench event in a large field magnet in a DEMO reactor

The superconducting magnet systems of future fusion reactors, such as a demonstration power plant (DEMO), will produce magnetic field energies in the 10s of GJ range. The release of this energy during a fault condition could produce arcs that can damage the magnets of these systems. The public safety consequences of such events must be explored for a DEMO reactor because the magnets are located near the DEMO's primary radioactive confinement barrier, the reactor's vacuum vessel (VV). Great care will be taken in the design of DEMO's magnet systems to detect and provide a rapid field energy dump to avoid any accidents conditions. During an event when a fault condition proceeds undetected, the potential of producing melting of the magnet exists. If molten material from the magnet impinges on the walls of the VV, these walls could fail, resulting in a pathway for release of radioactive material from the VV. A model is under development at Idaho National Laboratory (INL) called MAGARC to investigate the consequences of this accident in a large toroidal field (TF) coil. Recent improvements to this model are described in this paper, along with predictions for a DEMO relevant event in a toroidal field magnet.

First Operational Results with the SST-1 Superconducting Magnet & its Cryogenics

The Steady State Superconducting Tokamak (SST-1) at the Institute for Plasma Research was commissioned in 2013 with the successful experimental validations of its magnets and cryogenic systems. Subsequently, the first plasma in SST-1 has been obtained on June 20, 2013. Since then, the cryo-magnetic systems of SST-1 have been operating reliably, in successive plasma campaigns. Unlike other contemporary cable-in-conduit-conductor (CICC) based superconducting Tokamak magnets, SST-1 CICC in Toroidal Field (TF) winding packs are cooled with two phase helium from a dedicated 1.3kW helium refrigerator liquefier plant. The TF magnets in SST-1 together with other magnets have been supporting creditable plasma operational scenarios since then, with the TF magnets being charged up to a maximum of 2.0 T on the plasma major radius so far. The vapour cooled current leads of SST-1 are also uniquely operated with cooled helium vapour all along, instead of liquid helium being stored in the lower superconducting sections. The operational experiences with SST-1 TF magnets, PF magnets, and the supporting 5K and 80K cryogenic systems contributing towards successful operations of SST-1 plasma experiments are elaborated in this paper.

Study on mitigation of pulsed heat load for ITER cryogenic system

One of the key requirements for ITER cryogenic system is the mitigation of the pulsed heat load deposited in the magnet system due to magnetic field variation and pulsed DT neutron production. As one of the control strategies, bypass valves of Toroidal Field (TF) case helium loop would be adjusted to mitigate the pulsed heat load to the LHe plant. A quasi-3D time-dependent thermal–hydraulic analysis of the TF winding packs and TF case has been performed to study the behaviors of TF magnets during the reference plasma scenario with the pulses of 400s burn and repetition time of 1800s. The model is based on a 1D helium flow and quasi-3D solid heat conduction model. The whole TF magnet is simulated taking into account thermal conduction between winding pack and case which are cooled separately. The heat loads are given as input information, which include AC losses in the conductor, eddy current losses in the structure, thermal radiation, thermal conduction and nuclear heating. The simulation results indicate that the temperature variation of TF magnet stays within the allowable range when the smooth control strategy is active.

Quench characterization and thermo hydraulic analysis of SST-1 TF magnet busbar

Toroidal field (TF) magnet system of steady-state superconducting tokamak-1 (SST-1) has 16 superconducting coils. TF coils are cooled with forced flow supercritical helium at 0.4MPa, at 4.5K and operate at nominal current of 10,000A. Prior to TF magnet system assembly in SST-1 tokamak, each TF coil was tested individually in a test cryostat. During these tests, TF coil was connected to a pair of conventional helium vapor cooled current leads. The connecting busbar was made from the same base cable-in-conduit-conductor (CICC) of SST-1 superconducting magnet system. Quenches experimentally observed in the busbar sections of the single coil test setups have been analyzed in this paper. A steady state thermo hydraulic analysis of TF magnet busbar in actual SST-1 tokamak assembly has been done. The experimental observations of quench and results of relevant thermo hydraulic analyses have been used to predict the safe operation regime of TF magnet system busbar during actual SST-1 tokamak operational scenarios.

First Engineering Validation Results of SST-1 TF Magnets System

All SST-1 superconducting Toroidal Field (TF) and Poloidal Field (PF) magnets have been refurbished, integrated, and assembled onto the SST-1 machine shell in mid of 2012. Fabrication of sub nano-ohm, leak tight DC joints in superconducting magnets winding packs, enhancing the insulation strength in operating conditions, and test qualifying all the magnets prior to their integration in SST-1 machine shell were the primary refurbishment aspects. All assembled SST-1 TF magnets in SST-1 machine shell have since been successfully cooled down employing 1.3 kW SST-1 Helium Refrigerator/Liquefier system maintaining helium leak tightness under all temperature conditions. The TF magnets have experimentally demonstrated thermo-mechanical behavior as per the design as well as excellent flow uniformity amongst various parallel paths. Subsequently, these TF magnets have been increasingly charged so as to create a magnetic field of 1.5 T at the SST-1 major radius of 1.1 m. This is the first occasion, when cable-in-conduit wound toroidal field magnets have been successfully operating with two phase flow in a Tokamak application. The first plasma in SST-1 has been successfully obtained in June 2013 where the SST-1 TF magnets have demonstrated excellent functional characteristics. The confidence boosting engineering and functional validation test results of SST-1 TF magnets as well as its performance during the recent SST-1 plasma campaign have been elaborated in this paper.

Analysis of the Effects of the Nuclear Heat Load on the ITER TF Magnets Temperature Margin

In the International Thermonuclear Experimental Reactor, the nuclear radiation escaping from the vacuum vessel reaches the superconducting toroidal field (TF) magnets, affecting the temperature margin ΔTmar, that is, the difference between the current sharing temperature and the operating temperature. The TF magnets are designed to operate at a minimum margin ΔTmarmin = 0.7 K. Recent design activity on in-vessel components, for example, blanket, in-vessel coils for plasma stability, suggests a potential enhancement of the nuclear heat load, leading to a reduction of ΔTmar, which is accurately assessed in the paper using the validated 4C code. For the case when the margin goes below the minimum, different possible mitigation strategies are investigated: the first considers the possible reduction of the He bath temperature from the nominal 4.3 K down to 3.8 K, and is proven to be successful. The others consider the possible increase of the dwell time between plasma pulses, and is shown to be inadequate, or the decrease of plasma pulse duration, which turns out to be effective below 300 s.

Performance of Joints in SST-1 Magnets

Novel sub-nano ohm leak tight joints were one of the primary objectives of the refurbishment of Steady State Superconducting Tokamak (SST-1). In total, there are 130 such shake-hand type of joints which have been fabricated on the winding packs of SST-1 Toroidal Field (TF) and Poloidal Field (PF) superconducting magnets. These joints have been validated to be performing per their design specifications in the assembled SST-1 TF and PF magnets during the first cool-down and charging of the assembled and integrated SST-1 superconducting magnet systems. This paper describes the novel aspects of design, process establishments, technology development, testing and qualifications of such joints on the single magnets in both two phase and supercritical flow conditions as well as the joints performances in the assembled SST-1 TF and PF magnets.

Conceptual Design Study for a 16 Tesla Conductor Test Facility, SUCCEX

As a part of a conceptual design study for a steady-state demonstration reactor (K-DEMO) of which the major and minor radii are around 6.5 and 2.0 m, respectively, toroidal field magnets which can generate around 8 T at plasma center with a peak magnetic field of ~ 16 T are being studied. For conductor tests under such a high magnetic field, a new conductor test facility is required and a conceptual design activity is going on, in parallel. The conductor test facility, named SUCCEX (SUperConducting Conductor EXperiment), will feed the sample with a current up to 100 kA and the sample temperature will be varied from 4.5 K to ~ 20 K. Moreover, Sultan-like conductor tests, U-shaped joint-less sample tests will be conducted. The 16-T Nb3 Sn magnet, with a bore of ~ 1-m diameter, is divided into three concentric split pairs, inner coil (IC, peak field of 16.4 T), middle coil (MC, 13.6 T), and outer coil (OC, 7.9 T), connected in a series, with a nominal operating current of ~ 18 kA. Each split pair coil is made of its own distinctive rectangular shaped cable-in-conduit conductor (CICC). The overall design concept and preliminary results of electro-magnetic, structural, and thermo-hydraulic analysis are presented.

Analysis of Temperature Rise of TF Magnet During Plasma Discharges on EAST

For the safe operation of Experimental Advanced Superconducting Tokamak (EAST) with higher plasma performance discharge in future, it is important to estimate the operation state of Toroidal Field (TF) magnet. The outlet temperature rise of TF coils and TF cases measured in experiment is analyzed. It is found that the outlet temperature rise of TF coils and TF cases increases with plasma current. The outlet temperature rise of TF coils and TF cases induced by plasma disruption is a small value, about 0.01 K. In addition, the outlet temperature rise of TF coils is not related to the plasma duration, but the outlet temperature rise of TF cases seems to increase with plasma duration.

Completion of TF Strand Production and Progress of TF Conductor Manufacture for JT-60SA Project

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak, Europe, as part of its in-kind contribution within the Broader Approach agreement, will provide the toroidal field (TF) magnet system. For this purpose, Fusion for Energy is committed to procure about 29 km of TF conductor. The TF conductor is cable-in-conduit type and includes 486 strands (2/3 NbTi-1/3 copper) wrapped with a thin stainless steel foil and compacted into a rectangular stainless steel jacket. The procurement is split into two main contracts: one for TF strand manufacturing and the other for TF conductor cabling and jacketing. TF strand manufacture was completed while the TF conductor one is being finished. In the present paper, we draw an overview of both productions emphasizing on the quality control (QC) approaches and on aspects relevant to risk management of the JT-60SA tokamak operation. For the NbTi strand, the complete production overview is provided including extensive statistical considerations on NbTi strand critical performances ( IC, TCS). For the TF conductor, the overview also deals with collected results from acceptance tests and peripheral tests led for consolidating the QC (hydraulic tests, nondestructive examination, full-size sample cold tests in SULTAN facility).

Overview of Conductor Production for ITER Toroidal Field Magnet in Korea

The ITER toroidal field (TF) conductor is made up of superconducting Nb3Sn and copper strands assembled into a multistage, rope-type cable inserted into a conduit of butt-welded stainless steel jacket sections. For the ITER Project, the Korean Domestic Agency (KODA) took the responsibility of the procurement of 27 superconducting conductors for the ITER TF magnets. After concluding the Procurement Arrangement (PA) with the ITER International Organization in May 2008, KODA has been implementing the PA through four major industrial contracts: (1) Nb3Sn strand, (2) cable, (3) stainless steel jacket sections, and (4) jacketing. Prior to the production of conductors required for the TF coils, one 760-m-long copper dummy conductor and one 100-m-long superconducting conductor were fabricated for manufacturing process qualification. As of June 2013, 16 TF conductors were successfully manufactured. The full-size conductor performance tests in the SULTAN facility yielded very high performance. This article describes the technical requirements of the TF conductor and how KODA has been manufacturing the conductors with a high-level quality assurance/quality control system. It also presents the results of acceptance tests, including those of the SULTAN test.

Demountable Toroidal Field Magnets for Use in a Compact Modular Fusion Reactor

A concept of demountable toroidal field magnets for a compact fusion reactor is discussed. The magnets generate a magnetic field of 9.2 T on axis, in a 3.3 m major radius tokamak. Subcooled YBCO conductors have a critical current density adequate to provide this large magnetic field, while operating at 20 K reduces thermodynamic cooling cost of the resistive electrical joints. Demountable magnets allow for vertical replacement and maintenance of internal components, potentially reducing cost and time of maintenance when compared to traditional sector maintenance. Preliminary measurements of contact resistance of a demountable YBCO electrical joint between are presented.

Toroidal Field Analysis of the J-TEXT Tokamak

The toroidal field (TF) magnet of the Joint TEXT (J-TEXT) tokamak consists of 16 D-shaped copper coils interlinked in series. In this study, the finite element simulation software ANSYS is used to calculate the toroidal magnetic field and its ripple over the plasma region. Moreover, the possible error field caused by the busbar and the mounting error or noncoaxiality among the TF coils are also calculated. Results show that the busbar lead-in and the possible outward radial shift of a single TF coil will cause a relatively large error field.