International Thermonuclear Experimental Reactor Toroidal Field Coil Research Articles

Completion and Test of the First ITER TF Coil Winding Pack by Europe

The International Thermonuclear Experimental Reactor (ITER) magnetic system includes 18 toroidal field (TF) coils constructed using Nb3Sn cable-in-conduit superconductor. Each TF coil comprises a winding pack (WP) composed of seven double pancake modules stacked together, impregnated and inserted into a stainless steel coil case. Ten TF coils are being produced in Europe, under the responsibility of Fusion for Energy (F4E, the European Domestic Agency), while the remaining nine TF coils are being produced in Japan. F4E has implemented a strategy dividing the procurement into three packages. One is related to the construction of 70 radial plates (RP), another to the fabrication of 10 WP, and a third to the cold test and coil-case insertion of 10 WP. After 7 years of R&D and qualification activities and of industrial production, the first ITER TF coil WP has been completed in Europe. Factory acceptance tests, including leak, dimensional, and electrical tests at room temperature, were completed in May 2017 and the series production of the remaining nine TF WPs in Europe is underway. The first package has been completed and all 70 RP have been delivered. Commissioning of major tooling for the third package is to be performed at the end of 2017. In this paper, we report on the test of the first TF WP and on the status of the remaining production.

Flow balancing orifice for ITER toroidal field coil

Flow balancing orifices (FBOs) are used in in International thermonuclear experimental reactor (ITER) Toroidal Field coil to uniform flow rate of cooling gas in the side double pancakes which have a different conductor length: 99 m and 305 m, respectively. FBOs consist of straight parts, elbows produced from a 316L stainless steel tube 21.34 x 2.11 mm and orifices made from a 316L stainless steel rod. Each of right and left FBOs contains 6 orifices, straight FBOs contain 4 and 6 orifices. Before manufacturing of qualification samples D.V. Efremov Institute of Electrophysical Apparatus (JSC NIIEFA) proposed to ITER a new approach to provide the seamless connection between a tube and a plate therefore the most critical weld between the orifice with 1 mm thickness and the tube removed from the FBOs final design. The proposed orifice diameter is three times less than the minimum requirement of the ISO 5167, therefore it was tasked to define accuracy of calculation flow characteristics at room temperature and compare with the experimental data. In 2015 the qualification samples of flow balancing orifices were produced and tested. The results of experimental data showed that the deviation of calculated data is less than 7%. Based on this result and other tests ITER approved the design of FBOs, which made it possible to start the serial production. In 2016 JSC NIIEFA delivered 50 FBOs to ITER, i.e. 24 left side, 24 right side and 2 straight FBOs. In order to define the quality of FBOs the test facility in JSC NIIEFA was prepared. The helium tightness test at 10-9 m3·Pa/s the pressure up to 3 MPa, flow rate measuring at the various pressure drops, the non-destructive tests of orifices and weld seams (ISO 5817, class B) were conducted. Other tests such as check dimensions and thermo cycling 300 - 80 - 300 K also were carried out for each FBO.

Development of laser welding technology for fully austenitic stainless steel

The radial stainless steel plates (RPs) used for Toroidal field (TF) coils in the International Thermonuclear Experimental Reactor (ITER) are 13 m long, 9 m wide and 10 cm thick, which are quite large. Even though they are very large structures, high manufacturing tolerances and high mechanical strength at 4 K are required. It is also required that each RP should be fabricated every three weeks. Therefore, the authors intend to develop efficient manufacturing methods for an ITER TF coil RP. Laser welding is then selected as a welding method for RP. Especially, the development of high technology laser welding is necessary to prevent hot cracking in the material used for the RP; namely, fully austenitic stainless steel with high nitrogen content. The authors carried out trial laser welding experiments aiming at its application to RP. As a result, it was effective to make the angle of back inclination of the weld head at a uniform welding speed. It also seemed that the sensitivity of hot cracking could be reduced by optimizing the chemical compositions of material used for RP. The base material and the welded joints satisfied mechanical properties in 4 K. The application of the laser welding technology to the fully austenitic stainless steel was therefore demonstrated.

Progress in Procurement of ITER Toroidal Field Coil in Japan

The Japan Atomic Energy Agency (JAEA) has a responsibility to procure nine International Thermonuclear Experimental Reactor (ITER) toroidal field (TF) coils and 19 TF coil cases. The JAEA started sub- and full-scale trials to qualify and optimize the manufacturing procedures of TF coils and TF coil cases from August 2012. The major outcomes of these trials are development of an automatic winding system with high accuracy in conductor length measurement ±0.01%, evaluation of the conductor elongation of about 0.06% due to heat treatment, demonstration of the manufacturability of a full-scale radial plate with high accuracy, and establishment of the TF coil case manufacturing procedure satisfying tight tolerances. The JAEA thus started the series production of TF coils from the end of 2013 and TF coil cases from the beginning of 2014. Until now, 11 double-pancake (DP) windings were completed, and seven DPs were heat treated with elongation of the conductor from 0.063% to 0.077%, which are well consistent with the expected values. In addition, three main bodies of the basic segments of TF coil cases are being assembled. These successful achievements indicate good progress in the procurement of ITER TF coils.

Development of manufacturing technology of radial plate in superconducting coil for fusion reactor by diffusion bonding by hot isostatic pressing

The radial plates (RPs), which are used in the toroidal field (TF) coil in the International Thermonuclear Experimental Reactor (ITER), are quite large, 13 m tall and 9 m wide, but thin, 10 cm thick, and are made of stainless steel. Even though they are very large structures, they require very high manufacturing tolerances and high mechanical strength at 4 K. There will be similar requirements in the next-generation fusion reactor. Therefore, the authors intend to develop efficient manufacturing methods in parallel with ITER TF coil RP manufacture. The authors therefore performed trial manufacture of the RP segments using a diffusion bonding method, namely hot isostatic pressing (HIP). As a result of trials, it was clarified that even when HIP is applied, the mechanical characteristics of the base metal does not deteriorate. The machining period can be reduced by about 1/3 compared with the traditional manufacturing method. On the other hand, mechanical strength at 4 K is degraded due to weak bonding, that is no grain growth through joint, by HIP. However, an additional test indicates the promising possibility of a much better joint by higher temperature and joint surface-treated HIP. These results justified that RP segment manufacturing is not only possible, but it is a technically valid manufacturing method that satisfies all requirements.

Tensile test of SS 316LN jacket with different conditions

316LN stainless steel is selected as a material for Toroidal-field (TF) conductor jacket of International Thermonuclear Experimental Reactor (ITER). When energized, the ITER TF coils experience large pulsed electromagnetic forces that the conductor jacket itself must withstand. The conductor will be compacted, spooled and aged at approximately 650°C during manufacture. Therefore, the sample jackets were prepared under compaction, stretching and annealing to simulate the manufacturing process and operation of TF coils. The present Chinese TF jacket has good performance under standard conditions. In order to investigate more mechanical properties of 316LN jacket, the different cold working and annealing were applied to the raw materials. The samples were measured at 4.2K, 77K and 300K. Youngs modulus, yield strength (0.2% offset), elongation at failure and SEM images are reported. There is no big deviation among different conditions. The test results show that Chinese TF jacket has good performance.

Mitigation of the Temperature Margin Reduction due to the Nuclear Radiation on the ITER TF Coils

In the International Thermonuclear Experimental Reactor (ITER), the nuclear radiation escaping from the vacuum vessel reaches the superconducting toroidal field (TF) magnets, contributing to the reduction of the temperature margin ΔTmar, i.e., of the difference between the current sharing temperature and the operating temperature. The TF magnets are designed to operate at a minimum ΔTmar of 0.7 K. However, recent design activity on in-vessel components, e.g., the blanket and the in-vessel coils for plasma stability, suggests a potential enhancement of the nuclear heat load on the TF coils. A detailed reassessment of ΔTmar in the TF magnets during plasma operation is thus required, together with the study of possible strategies to mitigate the corresponding margin reduction. The extensively validated Cryogenic Circuit Conductor and Coil (4C) code, developed for the thermal-hydraulic analysis of transients in superconducting magnets, including their cooling circuit, is used here for the simulation of the standard ITER operating scenario. Among the different possible mitigation strategies, we consider the variation of the cooling He mass flow rate, both in the winding and, separately, in the casing, and its effectiveness in terms of ΔTmar is assessed.

Development of optimum manufacturing technologies of radial plates for the ITER toroidal field coils

Japan Atomic Energy Agency is studying rational manufacturing method and developing the optimum manufacturing technologies of the radial plates used in the toroidal field coils for the International Thermonuclear Experimental Reactor (ITER) in collaboration with the Japanese industries. Three sector form pieces were cut by plasma cutting machine from a hot rolled plate without any difficulties and one of them was machined to a 1.32-m long curved segment of the radial plate having the same size as the actual one. However, unacceptable large deformation about 5 mm flatness, which was not observed in 1-m long straight radial plate, was found after intermediate machining. Since it would be caused by groove direction against the hot rolled direction and/or curved shape of grooves, two trial manufactures of 0.4-m long straight radial plates have been performed to clarify the cause of the large deformation. Detailed investigation showed that the large deformation could be avoided if the groove direction would have been parallel to a rolling direction of the plate. Welding trials by using fiber laser technique was also performed and penetration of 15 mm could be obtained in a welding speed of 0.1 m/min at 5 kW laser power. An optimum manufacturing method has been proposed based on the development of manufacturing technologies.

Influence of bending strain on the critical properties of jacketed Nb3Sn strands for ITER

The tests analysis of the International Thermonuclear Experimental Reactor (ITER) model coils showed lower performances than expected. For that reason, a specific program has been launched in Europe for investigating the influence of bending strain on the transport properties of high-performances Nb3Sn strands inside cable-in-conduit conductors when experiencing significant Lorentz forces. To this purpose, an experimental campaign has been initiated on single Nb3Sn strands jacketed inside a 0.2 mm thick stainless-steel (SS) tube in order to be tested in mechanical ITER Toroidal Field (TF) Coils relevant conditions. In this article we describe in a first part the procedure investigated at CEA to impose a controlled pure bending strain upon jacketed stands and we present the results of the qualification tests carried out in the ENEA facility to assess the reliability of the chosen method. In a second part we propose a new model to simulate the strands behaviour experiencing bending strain together with an initial compressive strain. To understand our results, a key parameter considered is a specific weighting function relying on strand interfilament current redistribution capacity. At the end, this model enables us to explain more reliably the experimental results but also shows discrepancies which requires further improvements and extra refinements

Manufacturing Study and Trial Fabrication of Radial Plate for ITER Toroidal Field Coil

Japan Atomic Energy Research Institute (JAERI) has studied manufacturing processes of radial plates used in Toroidal Field coils of the International Thermonuclear Experimental Reactor (ITER). A manufacturing procedure, in which grooved radial plate blocks are connected by laser welding, was considered as the most rational one. A trial fabrication of a radial plate which has the same dimensions of the actual ITER was performed to clarify the feasibility of this procedure through the measurements of deformation during machining and welding. As a result, it was concluded that this manufacturing procedure could be used for the ITER because it could achieve the required tolerances for manufacturing of the radial plates

How should we test the ITER TF coils?

The international thermonuclear experimental reactor (ITER) toroidal field (TF) magnet system consists of 18 superconducting coils using a 68 kA Nb 3Sn conductor. In order to guarantee the performances of these coils prior to their installation, the test of at least one prototype coil at liquid helium temperature and full current is required. The test of all coils in the two-coil test configuration, with successive charging of each coil to nominal current is recommended. This requires a large test facility.

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.

Development and manufacturing of superconducting cable-in-conduit conductor for ITER

Russia as a member of Cooperation in the Engineering Design Activities for International Thermonuclear Experimental Reactor (ITER) takes part in manufacturing of superconducting cable-in-conduit conductor (CICC) for ITER toroidal field (TF) coils. JSC VNIIKP is responsible for Cr-plating, cabling and jacketing of TF conductor and develops complex technology for TF CICC production in RFHT. The technology consists of chrome coating, cabling (with compaction and wrapping) and jacketing (with insertion and compaction) operations and quality assurance (QA) procedure performance at every operation and on the finished conductor. There is a technological shop for plating of superconducting and copper wires with metal barrier plates. The shop includes the lines for surface treatment and galvanic chrome-coating, the autonomous system for water supply and regeneration of polluted water. The shop can produce about 100 km Cr-coated strands per week. Large park of tubular and planetary stranders, wrapping machine, compaction devices and other cabling equipment is used for production of sub-cables and cable of CICC. The jacketing line was designed and mounted in a facility 35 km south from Moscow during 1994-1996. The main technological and QA equipment is created, installed and tested. The jacketing line is fitted out with a take-up-pay-off unit, welding equipment, caterpillar, winch, rollgang (up to 1000 m), tube rolling mill, different tools and devices. The main steps of jacketing are assembling of conduit, control of welding joints, pulling cable through the conduit, compaction and conductor quality control. The description of the equipment and results of the trials are presented below.

Effect of a transient helium flow behavior on velocity of normal zone propagation in forced cooled fusion S/C magnets (for ITER)

A numerical analysis of normal zone propagation along the length of a cable-in-conduit type conductor for large scale magnets is described. The transient temperature propagation along the cable length, the helium pressure rise, and induced helium flow velocities in long cable cooling paths have been calculated on the basis of the computer code developed. As an example, Nb/sub 3/Sn conductors for the International Thermonuclear Experimental Reactor (ITER) have been considered. Results of this investigation show that during quench the velocity of the normal zone in the cable-in-conduit conductor can vary in a wide range from 1 m/s to 40 m/s. It is noted that these velocities depend greatly on the character of magnetic field variation along the conductor length. As a result, the maximum velocity of the normal zone in the ITER central solenoid is higher than that in the ITER toroidal field coil.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

SIMS-2 facility for ITER model toroidal field coils tests

The existing SIMS facility built in the Kurchatov Institute for testing the Tokamak T-15 toroidal coils is being reconstructed into an upgrading SIMS-2 facility for the ITER (International Thermonuclear Experimental Reactor) model pancakes. This facility makes it possible to investigate critical current, stability, and energy losses in the conductors under conditions close to that of the ITER toroidal field coils. The SIMS-2 consists of a superconducting magnet system SMS-12, a cryostat, a power supply, and data acquisition and cryogenic systems. The SMS-12 includes a background field solenoid producing a 12 T field inside a solenoid bore of 0.59 m ID, two (axial and radial) magnet presses creating transverse compression on the model pancakes, and pulsed coils simulating plasma disruption and poloidal field effects on the conductor. >