ITER Superconducting Magnet System Research Articles

A simple study for an optimized operation of the ITER Cryodistribution cold rotating machines

The ITER superconducting (SC) magnet system is composed of 4 distinct units: the Central Solenoid (CS) coils, the Toroidal Field (TF) coils, the coil encasing and supporting Structures (ST), and the Poloidal Field and Correction Coils (PF&CC or simply PF). Including another cold component unit, which is the Cryopump (CP) system, each unit is assigned its own Auxiliary Cold Box (ACB) with cold rotating machines (CRMs) to ensure efficient operation from a cryogenic perspective. The CRMs, which are the cold circulator (CCr) and cold compressor (CCp), are respectively in charge of generating the high mass flow rate (2.0–2.7 kg/s) and low temperature (3.7–4.2 K) of supercritical helium (SHe) that is supplied to each magnet unit and CP system. To maintain the proper cryogenic conditions of the magnets and CPs during fusion experiments, the ITER Cryogenic System can produce an average cooling power of 75 kW at 4.5 K, including the need for the 50 K gaseous helium (GHe) cooled high-Tc SC current leads. Depending on the operation mode of the ITER Tokamak, up to 28 kW of that power can be consumed by the CRMs due to compression of helium.In this study, we will explore possibilities for optimized operation of the CRMs by adjusting the mass flow rate and temperature of the SHe to minimize the heat of compression while maintaining the thermal stability of the ITER SC magnets and CP. By doing so, the CRM-heat-of-compression at 4.5 K can be reduced by more than 3 kW, providing additional “cryogenic” margins for higher-heat-load plasma experiments.

Voltages in the ITER magnets during operation

Powerful electromagnetic transients characterize the operation of the ITER superconducting magnet system. This is especially the case for so-called plasma disruptions, but also during plasma current initiation and termination and during the fast discharge of the magnets for quench protection. To model these electromagnetic transients the CARIDDI code (Albanese and Rubinacci, 1988) has been used to calculate the induced voltages in several magnet quench detection loops during operational events of interest. The numerical model implemented in CARIDDI, based on an integral formulation discretized in terms of edge elements, is shown to be particularly well suited for analysing these transients which require a high level of precision. Note that, although we will mostly present results obtained with CARIDDI, all calculations were cross-checked with other codes and methods, most notably (ANSYS Maxwell, 2022).

The Scaling Parameterization of ITER Superconducting Nb–Ti Strands Throughout Worldwide Production

The ITER superconducting magnet system will require approximately 650 tons of toroidal field and central solenoid Nb3Sn strands with different designs, and more than 250 tons of Nb–Ti strands. This called for a significant scale up of the worldwide production of Nb3Sn and Nb–Ti strands. Over the worldwide and mass productions, it is essential to accurately analyze and characterize the properties of ITER superconducting strands in terms of critical current $(I_{c})$ and ac loss, and thus determine the operational limits of the conductors, ultimately optimizing the operating scenarios of the Tokamak. In this paper, the proper scaling parametrization $I_{c}(B,T)$ and $n$ -value as a function of $I_{c} (n(I_{c}))$ are investigated for Nb–Ti strands throughout massive production. Despite the differences in ITER Nb–Ti strand's architecture and their composition, optimized Nb–Ti strand scaling parameterizations throughout production for each supplier have been reached with the minimal deviation ( $B, T$ window and even <5% in the operation condition region. This is to be applied for the analysis of the Nb–Ti conductors made from these stands and to assess possible evolution over production.

Optimization of the ITER Cryodistribution for an Efficient Cooling of the Magnet System

The ITER superconducting (SC) magnet system, which consists of central solenoid coils, toroidal field (TF) coils, TF structures, and poloidal field and correction coils, is cooled by supercritical helium (SHe) circuits located in dedicated auxiliary cold boxes (ACBs) of the cryodistribution (CD). Due to the increase in the nuclear heat load during the deuterium-tritium plasma phase, the necessary cooling power to maintain the thermal stability of the magnet system can be beyond the presently designed and contracted specification of the cryoplant (LHe and liquid nitrogen plants). Increasing the cryoplant capacity or performance in turn will significantly increase the project cost apart from the impacts on the utility infrastructures (cooling water, electricity, civil works, etc.). In this paper, we will present arrangements within the CD for an efficient and flexible operation. Instead of a common control, which is the present design, by individually controlling the LHe bath temperature of the ACBs, the cooling power can be concentrated to the magnet system in need (reduction in the SHe circuit temperature). In addition to this, a proposal to minimize the heat of compression of the cold rotating machines in order to further allocate the cooling power to the SC magnet system will be introduced.

Structural Analysis of ITER Toroidal Field Coil In-Cryostat Feeders

The ITER superconducting magnet system consists of 18 toroidal field (TF) coils, 6 poloidal field coils, 6 central solenoid modules, and 18 correction coils. Each coil is connected with a magnet feeder to support and control the high-current operation at cryogenic temperature. This paper presents the structural designs and finite-element (FE) analyses of the special construction of the TF in-cryostat feeders (ICFs). A global finite element model (FEM) was built to assess the distribution of the mechanical loads affected by gravitational, thermomechanical, and electromagnetic forces, as well as the displacement by the coils. Design challenges have been taken into account, and critical issues have been identified by global analysis. Valuable changes have been made based on global analysis results to solve the issues. Displacements obtained from global analysis were applied to local models as boundary conditions. Detailed FE analyses were performed to identify the local stress level in welded connections and the interaction between insulated bus bars and their clamps. The results of global and local analyses verify that the ICF ring box and bus bar jacket of the improved TF ICF design can withstand the operational loads with sufficient safety margin.

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.

An integrated model for the assessment of unmitigated fault events in ITER's superconducting magnets

A large amount of energy is stored in ITER superconducting magnet system. Faults which initiate a discharge are typically mitigated to quickly transfer away the stored magnetic energy for dissipation through a bank of resistors. In an extreme unlikely occurrence, an unmitigated fault event represents a potentially severe discharge of energy into the coils and the surrounding structure.A new simulation tool has been developed for the detailed study of these unmitigated fault events. The tool integrates: the propagation of multiple quench fronts initiated by an initial fault or by subsequent coil heating; the 3D convection and conduction of heat through the magnet structure; the 3D conduction of current and Ohmic heating both along the conductor and via alternate pathways generated by arcing or material melt. Arcs linking broken sections of conductor or separate turns are simulated with a new unconstrained arc model to balance electrical current paths and heat generation within the arc column in the multi-physics model. The influence under the high Lorenz forces present is taken into account. Simulation results for an unmitigated fault in a poloidal field coil are presented.

High Temperature Baking Test and Study on Axial Insulation Breaks

Axial insulation breaks are needed in ITER superconducting magnet system, which are used for separation of high voltage area from grounded cooling pipes system. To determine the maximum safe temperature that the insulating break can withstand without damage, such as preventing damage by overheating during welding of the insulating breaks to the helium cooling pipes for the superconducting magnets and high temperature baking to remove moisture, the glass transition temperature was tested by using the dynamic mechanical analysis with the standard sample made from epoxy resin. Furthermore, the high temperature baking test of axial insulation break was performed, further helium tight test and high voltage tests indicate the baked insulation break is in good condition.

Test Results of Pressure Head Mitigation of Supercritical Helium Across Cold Circulators at KSTAR for the Justification of the ITER Central Solenoid Cooling Circuit Design

The pressure head necessary for the forced convection of supercritical helium (SHe) across the ITER superconducting (SC) magnet system is generated by centrifugal-compressor type cold circulators. From the viewpoint of the cryogenic system, simulation results indicate that the main challenge of the ITER Central Solenoid (CS) operation is the peaking pressure head of SHe across the cold circulator induced during the rapid ramping of electrical current on the coils. Excess variation in pressure head will push the operation point of the rotating machine beyond the surge line in the compressor map which can trip the circulator and in the worst case result in damages. Therefore, in order to avoid the occurrence of a surge, it has been proposed to open adaptively a cryogenic valve, installed in parallel with the CS cooling channels, during the high pressure head instances. Such an action creates an additional flow channel which consequently suppresses the pressure head peak. The proposed strategy has been numerically benchmarked and tests have been performed in existing cryogenic facilities. In this paper, we will present the test results of pressure head mitigation across the cold circulator dedicated to the KSTAR CS and Poloidal Field (PF) coil cooling circuit with the main purpose of demonstrating the safe operation of the ITER CS circulator as well as cooling circuit during plasma shots. The tests were performed by progressively increasing the coil by-pass valve opening during identical PF/CS current shots and monitoring the tendency of the maximum pressure head variation across the cold circulator. Also, preliminary results of pressure head mitigation during real plasma experiments will be introduced.

Tryout of the vacuum impregnation procedure for the ITER PF1 coil on the VPI mold made of plastic shell

The PF1 Coil as part of the ITER superconducting magnet system is intended for positioning and shaping of the magnetic poloidal field. The technical specification for the PF1 coil requires a full-scale simulation of the basic technological processes including vacuum pressure impregnation (VPI). Usually, a VPI mold is made of massive stainless steel walls to get the required quality of the monolithic structure of the pancake insulation by hot-curing compound. As a result, we obtain a high-cost furnace with a considerable amount of steel inside to be heated up to the specified temperature; excessive energy consumption for the furnace heating system; problems with the vacuum tightness of the VPI mold. The impregnation procedure using a VPI "plastic" mold was proposed so as to avoid the above mentioned shortcomings associated with the use of the furnace made of stainless steel.

Quench Detection in ITER Superconducting Magnet Systems

The quench of one of the ITER magnet systems is an irreversible transition of the conductor from superconducting to normal resistive state. The normal zone propagates along the cable-in-conduit conductor, dissipating a large power. The detection has to be fast enough (1 to 2 s) to initiate the dumping of the magnetic energy and avoid irreversible damage of the systems.The experience of CEA is based on the operation of the superconducting tokamak Tore Supra for more than 20 years. In support of ITER, CEA was also very involved in quench detection investigations during these past 3 years.The primary quench detection in ITER is based on voltage detection, the most rapid detection. The very magnetically disturbed environment during a plasma scenario makes the voltage detection particularly difficult, inducing large inductive components across the pulsed coils (10 kV) or coil subcomponents. Voltage compensations therefore have to be designed to discriminate the resistive voltage associated with the quench.A secondary detection based on a thermohydraulic signals system also has to be investigated to protect the environment in case of a nondetected quench, especially for the largest ITER system, which is the toroidal field system with a stored energy of 40 GJ.

Progress of ITER Superconducting Magnet Procurement

Abstract The ITER superconducting magnet system consists of 18 Toroidal Field (TF) coils, 1 Central Solenoid (CS), 6 Poloidal Field (PF) coils and 18 Correction coils (CC). The TF conductors will be manufactured by China (7%), EU (20%), Korea (20%), Japan (25%), Russia (20%) and US (8%), TF coils by EU (10 coils) and Japan (9 coils), in which one spare is included, all TF coil cases by Japan, all CS conductors by Japan, all CS (7 modules including a spare), PF conductor by China (65%), EU (21%) and Russia (14%), PF coils by EU (5 coils) and Russia (1 coil), all CCs by China and all feeder by China, respectively. Since the TF coil manufacture is one of long-lead items, procurement of the TF conductors have been started. More than 40 TF conductors have already been fabricated. Large-scale trials for TF coil manufacture have also been started and successful results were obtained in both EU and Japan, such as manufacture of full-scale radial plates. The trials for PF coil and CC has been done by Russia and China.

Thermo-Mechanical Instrumentation of the ITER Magnet Structures

The ITER superconducting magnet system (Toroidal Field coils (TF), Central Solenoid (CS), Poloidal Field coils (PF), Correction Coils (CC) and Feeders) represents a total weight of approximately 10 000 tons. It is designed to withstand gravitational and seismic forces, stresses induced by thermal contractions during cool-down from 300 K to 4.5 K, and large Lorentz forces in the superconducting coils. Although not part of the Safety Important Class (SIC) components, the sensors (stress, displacement, thermometer) used to monitor the thermo-mechanical behavior of the structures are key diagnostic means to assess the design, support operation and survey possible fatigue effects over the 20 years' lifetime of the Tokamak. The sensors and their wiring will be submitted to unique and severe environmental operating conditions; they will have to operate in the cryostat's high vacuum (<;10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">-4</sup> Pa total gas pressure) at magnetic inductions of several Tesla and at low temperature (4.5 K) in the presence of a neutron fluence on the order of 10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">22</sup> m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">-2</sup> - which implies a gamma dose higher than 10 MGy - for a period of 20 years. In addition, the fast and slow cycling of very large currents in the superconducting magnets and plasma will generate large eddy currents and, consequently, heat loads and electromotive forces (voltages) in the various parts of the structures, as well as a large electromagnetic noise. This paper presents the design philosophy and the chosen sensor technologies, some of which require further development in collaboration with the industry so that they can meet the objectives and the severe ITER operating conditions. In particular, only 20% of the near 1 000 measuring points for the thermo-mechanical data of the ITER magnet structures will rely on “classical” copper-wire technologies whereas 80% will rely on specifically developed optical-fiber-based sensors.

Key Components of the ITER Magnet Feeders

Now that ITER is entering construction, many of its systems are in the final stages of design and analysis. Among them the 31 feeders, which will be supplied in-kind by the Chinese ITER partner. The feeders supply the electrical power and cryogens through the warm-cold barrier to the ITER superconducting magnet systems. They are complex systems with their independent cryostats and thermal shields, densely packed with many components, such as the current feeds, the cryogenic valves and High-Voltage (HV) instrumentation hardware. Some of the feeder components are particularly critical and have been designed with great care. Among them the High-Temperature Superconductor (HTS) current leads, designed for unprecedented currents, the 30 kV class, Paschen-hard HV insulation and the bus bar support system, designed to react the multi-ton Lorentz-forces from the bus bars at minimal heat load. This paper discusses the design challenges for these (and other) key components of the ITER magnet feeders.

Status of the ITER Manufacturing Database for the Magnet System Procurement

All 21 Procurement Agreements (PAs) related to the ITER superconducting magnet system were signed in March 2011. Six Domestic Agencies (DAs) of the seven, namely China, EU, Japan, Russia, the United States and Korea have to deliver components of the magnet system like TF coils and Feeders. In order to organize the complicated international supply-chain, the ITER Manufacturing Database (MD) has been developed as a sub-project of the ITER Engineering Database (EDB) which aims to become the official Product Lifecycle Management (PLM) system of the ITER project. The first version of the PLM was released in August 2011. One of the advantages of the MD is to provide clear definitions of the work conditions and resultant deliverables to all stakeholders for a common understanding and better mutual communication. The work conditions (e.g. applicable code/standard) are basically agreed by all stakeholders after the preparation phase. They are then prepared in the MD as four customized User I/F (interface) screens: the interactive Parts List, the Manufacturing Inspection Plan (MIP), the Welding Quality Inspection and Production Plan (WQIPP) and the Tooling List. Users only upload the required reports (e.g. the mill certificate for each part, test reports) into the corresponding User I/F. Consequently, the filled up User I/F becomes a deliverable, the Manufacturing Dossier.

Quench Detection in the ITER Magnet System

The quench detection system in ITER superconducting magnet systems is a real challenge for the poloidal field (PF) system and the central solenoid (CS) due to their fast-pulsed operation. An existing CEA code allowing very precise magnetic field calculations in tokamaks (Traps) has been upgraded (TrapsAv) to calculate the inductive voltages induced across any coil sub-element during a plasma scenario. The envisaged primary detection system is based on voltage measurements. For the CS, a detection system is proposed and discussed based on the central difference average balance of 3 neighboring double pancakes. For the TF system, the presently considered compensation circuit in ITER is a strip co-wound on to the conductor. A simple analytical calculation is proposed to estimate the voltage during a plasma disruption or during the scenario across the co-wound strip and across the conductor as well, due to the toroidal component of the plasma current.

World-Wide Benchmarking of ITER ${\rm Nb}_{3}{\rm Sn}$ Strand Test Facilities

<para xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> The world-wide procurement of <formula formulatype="inline"><tex Notation="TeX">${\rm Nb}_{3}{\rm Sn}$</tex></formula> and NbTi for the ITER superconducting magnet systems will involve eight to ten strand suppliers from six Domestic Agencies (DAs) on three continents. To ensure accurate and consistent measurement of the physical and superconducting properties of the composite strand, a strand test facility benchmarking effort was initiated in August 2008. The objectives of this effort are to assess and improve the superconducting strand test and sample preparation technologies at each DA and supplier, in preparation for the more than ten thousand samples that will be tested during ITER procurement. The present benchmarking includes tests for critical current <formula formulatype="inline"> <tex Notation="TeX">$(I_{\rm c})$</tex></formula>, <formula formulatype="inline"> <tex Notation="TeX">$n$</tex></formula>-index, hysteresis loss <formula formulatype="inline"> <tex Notation="TeX">$(Q_{\rm hys})$</tex></formula>, residual resistivity ratio <formula formulatype="inline"><tex Notation="TeX">$(RRR)$</tex></formula>, strand diameter, Cu fraction, twist pitch, twist direction, and metal plating thickness (Cr or Ni). </para>

Test Results of 52/68 kA Trial HTS Current Leads for ITER

High Temperature Superconductor (HTS) current leads provide the transition of the current transfer system from room temperature to the 4.5 K. For the ITER superconducting magnet system, a total of 60 HTS current leads are required, with a total nominal current capacity of 2.64 MA. By the use of HTS leads significant power savings can be achieved during the operation of the ITER magnet system. Based on the functional specification by the ITER Organization (IO), ASIPP manufactured and tested a first pair of trial HTS current leads for the ITER coils. The cryogenic test of these 68 kA and 52 kA trial HTS leads was performed at ASIPP using the cryoplant and power supplies of the EAST Tokamak. The following paper presents the results of the tests and concludes with a discussion of the design features of the next generation prototype of the ITER HTS leads based on the results of the trial lead tests.

Technology development for the construction of the ITER superconducting magnet system

The Japanese contribution to the construction of the ITER superconducting magnet system is to procure 25% of the toroidal field (TF) conductors, about half of the TF coil winding packs, all of the TF coil structures and all of the central solenoid conductors. The Japan Atomic Energy Agency has started extensive technology development for the preparation of these procurements. These include (1) trial fabrication of the superconductors at industry level and their performance demonstration, (2) manufacturing studies on the TF coil winding pack and structures, (3) full scale trial fabrication of TF coil structures and (4) the establishment of a database on the structural materials. All these activities are performed in close collaboration with Japanese industry involving around 20 major companies. Fruitful results have so far been obtained, although further development of the technology is required in some areas, including the qualification of conductors and the establishment of welding procedures of structures.

Japanese contributions to the procurement of the ITER superconducting magnet

Japan is expected to make a major contribution to the procurement of the ITER superconducting magnet system, including the TF coil windings and structures, and conductors for the TF coil and CS. JAEA has started preparation activities for the ITER construction in tight collaboration with industries. These activities include manufacturing studies to identify detailed fabrication processes and tools for critical components, and manufacturing demonstrations at full scale level based on the studies. So far, high-performance Nb 3Sn strands that satisfy the ITER requirements have been developed and full scale trial fabrications of new cryogenic materials, such as JJ1 and strengthened 316LN, have successfully been performed for TF coil case and radial plates. Results from these activities will provide a firm technical basis to achieve required performance of the magnets while maintaining both project schedule and cost.