Measurements of stability margins and current distribution in large-scale Nb{sub 3}Sn cable-in-conduit conductors

The main goal of the ITER Conductor Testing Program is to validate the design and the fabrication process of the CICC (Cable-In-Conduit Conductor) and its joints, in full-size samples. For testing large-scale cable-in-conduit conductors, several practical experimental techniques have been developed to specifically address the issues of conductor stability. The following described techniques enable experimentalists to quantify the effects and thus provide useful tools to improve the conductor performance in magnet applications. In this paper, the authors, present two experimental techniques: (a) stability margin calibration and (b) current distribution. Both are much needed in the laboratory for testing the conductor stability

Analysis of the DC performance of the ITER CSI coil using the 4C code

The DC performance of the ITER Central Solenoid Insert (CSI) coil, a single layer solenoid wound using the same Nb3Sn conductor that will be adopted for the 3L module of ITER CS, was measured during the 2015 test campaign in different magnetic field and current operating conditions, before and after electromagnetic and thermal cycles, as well as before and after quench tests. The 4C thermal-hydraulic code is applied here to the analysis of the CSI performance: first, the free parameters of the model are calibrated; then, the model is validated against measurements not used for its calibration. The model is then used to compute the current sharing temperature, to be compared with the measured jacket temperature, and to assess the performance after quench tests

The quench recovery analysis of the JT-60SA superconducting magnets

JT-60SA is one of the experimental nuclear fusion reactors with superconducting magnets. It is a joint international research and development project involving Japan and Europe. The temperature distribution changes in recovery is investigated.The quench recovery period is necessary to be confirmed. Generally, the maximum temperature drop of magnets is able to be confirmed by checking the thermometer attached to the outlet of the helium flow path. However, the maximum temperature of the JT-60SA central solenoid (CS) is not able to be measured during quench recovery. The flowing paths of CS is C-shaped and both of the outlets and the inlets of helium are on the outer periphery surface of the CS modules. Due to this C-shaped flowing path, heat exchanges between the inlet flow paths and the outlet flow paths. The CS outer periphery side becomes colder than the inner periphery side. The typical issue is the CS inside temperature is not able to be measured by the thermometers on the flowing paths. In this work, the CS temperature distribution changes during quench recovery is calculated and the period necessary for recovery is investigated.A CS module is composed of the 52 layers pancake coils. The 26 helium flowing paths are in a one module. The refrigerator supplies helium at 4.4 K to each flowing paths in nominal operation. In case of a quench, the refrigerator stops helium supply in order to shut out large heat load from the quenched magnet. The temperature distribution of the quenched CS will be smoothed by a heat conduction between each pancake coils while helium is stopped. Helium will be supplied again when the magnet pressure become low enough.The temperature distribution changes are calculated by using the thermal fluid simulation codes

Design of JT-60SA Cryodistribution components

JT-60SA is a fusion experiment tokamak device using superconducting magnets to be built in Japan. This joint international project involves Japan and Europe. In this work, we presents the design of cryodistribution and its components which are composed of a main transfer line (TL) and valve boxes (VB).Five coolant loops are distributed between a helium refrigerator system (HRS) and cold components. Super critical pressure helium (SHe) of 4.5 K and 0.5 MPa supplied to 18 toroidal field coils, 6 equilibrium field coils and 4 central solenoid modules (LOOP1 & 2). SHe of 3.7 K and 0.5 MPa is supplied to divertor cryopumps (LOOP3). Gaseous helium (GHe) of 80 K and 1.4 MPa is supplied to radiation thermal shields (LOOP4). GHe of 50 K and 0.4 MPa is supplied to cold ends of high temperature superconducting current leads (LOOP5).TL is a vacuum heat-insulation multiple piping, of which the length is about 45 m, and connects between HRS and the tokamak cryostat. All 5 supply lines, 4 return lines and 2 control valves are installed in TL. The outer vacuum pipe diameter is 965.5 mm and the inner coolant pipe diameter are 108.3 mm for LOOP 1/2/4 and 59.0 mm for LOOP 3/5. A vacuum partition between HRS and the tokamak cryostat is located near the middle of TL in a longitudinal direction.VB contains cryogenic valves and measurement devices to control the cold helium flow. Eleven VBs are installed around the tokamak cryostat. Dimensions of VB body are 2 m in height and 1.4 m in diameter. Almost all cold helium lines from HRS are firstly into VBs through TL. Impulse lines, orifice plates, and resistor elements are installed at the pipes in VB for measurement of the pressure, the flow rate, and the temperature of coolant helium.第26回国際磁石技術会議（MT26 International Conference on Magnet Technology

Analysis of Quench Propagation in the ITER Central Solenoid Insert (CSI) Coil

The Central Solenoid Insert (CSI) coil, a single-layer Nb3Sn solenoid, wound using the same conductor of the 3L module of the ITER Central Solenoid, was tested in 2015 at the National Institutes for Quantum and Radiological Science and Technology (former JAEA) Naka, Japan, inside the bore of the Central Solenoid Model Coil. At the end of the test campaign, quench tests were carried out to study the quench initiation and propagation. Different delay times (up to 7 s) between quench detection and current dump were set, in order to explore to which extent the dump could be delayed without exceeding the maximum allowed hot spot temperature. The experimental data for different time delays are presented and compared, showing a good reproducibility of the measurements and confirming the safe operation of the coil during these tests. The previously developed and already extensively validated 4C thermal-hydraulic code is then used to model the transient up to the current dump and a comprehensive comparison between the simulation results and the measurements is presented, including the evolution of the local voltages and of the jacket temperature distribution along the conductor, the quench front propagation, and the SHe pressurization and mass flow rate behaviour measured at the CSI inlet and outlet. The good agreement between simulation results and measurements confirms the validation of the 4C code for this type of transients and the code is then used to explain the acceleration of the quench and to get an improved estimate of the hot spot temperature

Analysis of the DC performance of the ITER CSI coil using the 4C code

The DC performance of the ITER Central Solenoid Insert (CSI) coil, a single layer solenoid wound using the same Nb3Sn conductor that will be adopted for the 3L module of ITER CS, was measured during the 2015 test campaign in different magnetic field and current operating conditions, before and after electromagnetic and thermal cycles, as well as before and after quench tests. The 4C thermal-hydraulic code is applied here to the analysis of the CSI performance: first, the free parameters of the model are calibrated; then, the model is validated against measurements not used for its calibration. The model is then used to compute the current sharing temperature, to be compared with the measured jacket temperature, and to assess the performance after quench tests