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An ITER TF model coil conductor (DP4) was examined to determine electromagnetic and mechanical AC loss under transverse cyclic loading.

An evaluation of a special cryogenic press constructed to study the electrical and mechanical characteristics of full-size ITER conductor samples under transverse, mechanical loading conditions is presented. A superconducting dipole coil is used to provide the required magnetic field for magnetization measurements. The results, obtained after several tens of full loading cycles, indicate a significant reduction of the cable coupling loss time constants after a number of cycles.

Research was conducted to examine the mechanical and electrical characteristics of full-size ITER multistrand Nb3Sn cable-in-conduit conductor samples under transverse, mechanical loading. The transverse magnetic force that takes place when the conductors are used in a coil was simulated. A summary of findings derived with up to several tens of full loading cycles are presented.

Research was conducted to examine the mechanical and electrical properties of full-size ITER-CS cable samples under transverse, mechanical loading. Fatique tests were conducted including loading the conductor from zero up to 260 kN for a large number of cycles to investigate the evolution of n-tau adn R(sub c) under realistic ITER lifetime conditions. Results of the mechanical calculations and the electromagnetic losses and interstrand contact resistances are presented and discussed.

This paper describes the result of the ITER feeder main busbar joint sample qualification test as confirmation of the requirement of busbar joint resistance, 2 nΩ at 70 kA at zero background field, as well as that of joint performance in various magnetic fields to investigate stability and current distribution characteristics in feeder-type joint box. The results support the quality of the joint manufacturing process for ITER main busbar joint. The qualification sample design was prepared to be tested in SULTAN facility. The SULTAN joint sample consists of joints to be qualified at the level of the peak field and upper terminations. In bottom joints, twin-box feeder-type praying hands configuration is applied. In upper terminations, one of them is made with solderfilled cable for optimum current distribution. The other takes the same length of the copper sole and contact with the busbar cable as those positioned in bottom of the sample. The sample undergoes a test program, which includes joint resistance measurement, ac losses, and stability margin test. The outcomes of those test programs are reported. [ABSTRACT FROM AUTHOR]

The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Science (ASIPP) by incorporating the results of mechanical and thermal analyses as well as the system integration and assembly tolerances in the present CAD model. The feeder design is being finalized progressively, and will be delivered to the Chinese Domestic Agent (CNDA) for further procurement arrangement (PA) activities. Pre-PA manufacturing studies and tests performed at ASIPP have been effective in clarifying feeder design feasibility and component manufacturability. This paper reports the recent advancements on feeder design, analysis and manufacturing studies. [ABSTRACT FROM PUBLISHER]

The ITER magnet coils are wound from Cable-In-Conduit Conductors (CICC) made up of superconducting and copper strands assembled into a multistage, rope-type cable inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the Toroidal Field (TF) and Central Solenoid (CS) coils require about 500 tons of Nb3Sn strands while the Poloidal Field (PF) and Correction Coil (CC) conductors need around 250 tons of Nb-Ti strands. The required amount of Nb3Sn strands far exceeds pre-existing industrial capacity and calls for a significant worldwide production scale up. [ABSTRACT FROM PUBLISHER]

The Procurement of the feeder system for The International Thermonuclear Experimental Reactor (ITER) tokamak device started in ASIPP (China) in 2006. The relevant tasks of structural design and analysis have now come to an end, but the R&D activities started in 2009 for critical feeder components and key technology are still ongoing. Following the signature of the Procurement Arrangement between China and the ITER Organization (IO) in 2010, a great number of R&D tasks have been launched with the purpose of testing and verifying the original design, leading to design changes and optimizations. This paper mainly discusses the progress of the feeder design and R&D activities; it also presents some test results. [ABSTRACT FROM PUBLISHER]

This paper summarizes the results of the static structural analyses, which were conducted in support of the ITER magnet feeder design with the aim of validating certain components against the structural design criteria. While almost every feeder has unique features, they all share many common constructional elements and the same functional specifications. The analysis approach to assess the load conditions and stresses that have driven the design is equivalent for all feeders, except for particularities that needed to be modeled in each case. The mechanical analysis of the feeders follows the sub-modeling approach: the results of the global mechanical model of a feeder assembly are used as input for the detailed models of the feeder' sub-assemblies or single components. Examples of such approach, including the load conditions, stress assessment criteria and solutions for the most critical components, are discussed. It has been concluded that the feeder system is safe in the referential operation scenarios. [ABSTRACT FROM PUBLISHER]

As an important method of quality confirmation for the ITER Feeder system insulated components, high voltage test at Paschen condition need to be carried out once after these components are manufactured. The Paschen test is much more efficient for detecting the micro cracks or defects in solid multi-composite high polymer insulation materials than the conventional high voltage test done in atmospheric condition, and almost without any damage for the qualified components. [ABSTRACT FROM PUBLISHER]

Following the design, fabrication and test of a series of trial leads, designs of the three types of current leads required for ITER have been developed, and targeted trials of specific features are in progress on the way to fabrication and testing of prototype units. These leads are of the hybrid type with a cold section based on the use of high temperature superconductor (HTS) and a resistive section cooled by forced flow of helium gas, optimized for operation at 68 kA, 55 kA and 10 kA. The leads incorporate relevant features of the large series of current leads developed and constructed for the CERN-LHC, relevant features of the trial leads built for ITER, and additional features required to fully satisfy the exigent constraints of ITER with regard to cooling, insulation, and interfaces to feeder and powering systems. In this report a description of the design of the leads is presented, together with plans for the preparation of prototype manufacture and testing at ASIPP. [ABSTRACT FROM PUBLISHER]

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. [ABSTRACT FROM PUBLISHER]

The ITER Magnet system refid="ref1"/ consists of four main coils sub-systems: 18 Toroidal Field Coils (TF-coil), a Central Solenoid (CS), 6 Poloidal Field Coils (PF-coil) and 18 Correction Coils (EFCC). The main contract of the EFCCs supply is awarded to the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) by the Chinese Domestic Agency (CNDA). According to the pre-qualification program, ASIPP is implementing the procurement phase to qualify and validate key technologies and manufacturing methods. The Correction coils qualification activities are conducted within the framework of the procurement arrangement set up between the ITER Organization and CNDA. [ABSTRACT FROM PUBLISHER]

Recent design changes to the ITER Poloidal Field (PF) coils and PF conductor layout have been implemented to give a greater operational window for low li (self-inductance) plasmas during burn, extend the operating window for plasmas with currents above 15 MA, and improve the plasma vertical stability control. In addition, the PF and CS (Central Solenoid) operating window has been updated leading to higher currents and peak fields in some of the PF coils. Altogether, this results in higher heat generation in the PF conductors due to AC losses. To confirm that the range of heat loads is acceptable for the PF conductors, a time-dependent thermo-hydraulic analysis of the PF coils has been performed with a model based on the GANDALF code. The deposited heat due to AC losses in the conductors, thermal radiation, thermal conduction, and nuclear heating are given as input data. The results show a moderate impact on minimum temperature margin of the PF conductors during the baseline scenario, stemming from the design modifications. [ABSTRACT FROM AUTHOR]

The ITER Organization (10) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly developing the 68 kA current leads using High Temperature Superconductors for the superconducting Toroidal Field (TF) magnet system of the International Thermonuclear Experimental Reactor, ITER. The proposed design consists of a conventional helium cooled heat exchanger operating between 65 K and 320 K and an HTS module covering the low temperature end using Bi-2223 tapes. The first HTS current lead prototypes will be tested in the EAST tokamak facility at ASIPP. This paper discusses the design of the first, 68 kA, TF HTS lead, which is the main emphasis of this first stage of the ITER current lead development program. [ABSTRACT FROM AUTHOR]

ITER PFCI has been manufactured in the Europe and installed into the ITER Test Facility in Naka, Japan. The conductor is NbTi cable-in-conduit conductor with thick square stainless steel jacket and almost identical with the design of the ITER PF coils. The main objective of this test is the characterization of the conductor and joints at the conditions relevant to the ITER PF coil operation. Intermediate joint is located in the winding at relatively high field to examine its performance. The main items in the PFCI test program are thermo-hydraulic test, DC mode test, cyclic test and pulse mode test. The PFCI and CSMC were successfully cooled down to cryogenic temperature within 450 hours. The test of the PFCI was performed from May to August 2008. The key technology of the installation, the test methods and procedures, and some preliminary results of cool-down are described in this paper. [ABSTRACT FROM AUTHOR]

To investigate the conductor behavior during a quench, quench tests of the Central Solenoid Insert Coil (CSIC) in both DC and pulse modes were carried out with various initial conditions in the Central Solenoid Model Coil Facility. The Nb3Sn cable-in-conduit conductor with a similar design and configuration to the conductor for the ITER Central Solenoid (CS) was used. The inductive heater, attached on the conductor at the center of the length and the highest field region, initiated an artificial quench in the quench tests. A quench has also occurred during the pulse operation with the ramping rates of 0.4–2 T/s, in the ramp rate limitation tests of the CSIC. Simulations of electric, thermal and hydraulic behavior of the conductor during quench tests were carried out by using the ‘Gandalf’ thermohydraulic simulation code. The experimental results were compared with the simulation and good agreement was obtained. The implication for quench detection in ITER is also discussed and it is confirmed that the hot spot temperature of the ITER-CS is lower than the design criteria (<150 K). [ABSTRACT FROM AUTHOR]

At least two independent methods for quench detection should be used for the magnets of the ITER size, because of the potential severe impact of a failure. The voltage tape method and the flow meter method, which were used in the CS model coil (CSMC), are both considered for the ITER Central Solenoid (CS). The voltage tap method is primary due to its quick response. The ITER CS consists of six pancake wound modules, which are operated with individual operating current patterns in ac mode. The induced voltage in the windings must be compensated to detect the voltage due to any normal transition during the pulse operation. We have investigated the optimum configuration for pick-up coils (PC) for compensation. The results of simulations show that the compensated voltages are very low (70 mV) compared with the inductive voltage and adequate normal voltage sensitivity is obtained. The hot spot temperature in the CS during the operation was estimated from the simulation and the experimental measurements of the CSMC quench. The hot spot temperature estimated is 128-144 K, lower than the ITER design criterion (150 K). It is shown that the detection system using the PCs could be designed with a high enough detection sensitivity. [ABSTRACT FROM AUTHOR]