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Celebrity worship refers to individuals' social identification with and emotional attachment to their idols. In this study Experiment 1 explored participants' attentional bias toward idol faces during the attentional orienting phase of a dot-probe task, and Experiment 2 compared attentional bias for high- and low-worship groups during the attentional release phase using a dot-probe task and a spatial cueing task. Results showed that the high- (vs. low-) worship group exhibited shorter reaction times in the condition where the idol face and probe dot were consistent, and that their celebrity worship scores were positively correlated with their attention bias index. The high- (vs. low-) worship group also exhibited longer reaction times when idol faces were used as invalid stimuli. In conclusion, the high-worship group showed attentional facilitation and attentional disengagement difficulties for idol faces, and higher worship scores were associated with a stronger attentional bias. [ABSTRACT FROM AUTHOR]

High temperature superconducting current leads (HTS-CL) are designed to supply the current to the large superconducting ITER magnets for the operation with reduced heat load to the cryogenic system. The Toroidal field (TF) current leads are the largest with a current capacity of 68 kA each. The Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) is responsible for the supply of the current leads based on a design jointly developed with the ITER organization. Before the supply of the TF HTS-CL series, a pair of prototypes was manufactured in 2014 by ASIPP and associated manufacturers according to previously qualified manufacturing procedures. Rigorous quality control measures were developed and applied in preparation for series manufacturing. To verify compliance of the prototypes with the ITER specification, thorough testing was conducted in 2015. The test items under particular scrutiny were: the pressure drop in the counter-flow heat exchanger, the loss of flow accident test after steady state operation at 68 kA current, the so-called overheating time of the HTS module following an induced quench, the electrical resistances of the soldered joints inside the lead assembly (i.e., the low temperature superconducting (LTS) to busbar and the HTS to LTS joints), and the conduction heat load per lead to the 4.5  K end. In this paper, the main manufacturing steps are discussed, and test results are presented and discussed. [ABSTRACT FROM AUTHOR]

To carry current for the ITER correction coils, the 10-kA high-temperature superconducting current leads studied here are designed. Current leads provide the transition from 4.5-K low temperature to room temperature. This paper summarizes the major design features of the prototypes, which is followed by a discussion of the manufacturing and testing. The test results approved their excellent performance on low joint resistance and long loss-of-flow accident time. The overheating time, mass flow, and heat loads to 5-K ends also reached the expectation. [ABSTRACT FROM AUTHOR]

Three types of high temperature superconducting current leads (HTSCL) are designed to carry 68 kA, 55 kA or 10 kA to the ITER magnets. Before the supply of the HTS current lead series, the design and manufacturing process is qualified through mock-ups and prototypes. Seven mock-ups, representing the critical technologies of the current leads, were built and tested successfully in the Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) in 2013. After the qualification some design features of the HTS leads were updated. This paper summarizes the qualification through mock-ups. In 2014 ASIPP started the manufacture of the prototypes. The preparation and manufacturing process are also described. [ABSTRACT FROM AUTHOR]

ASIPP will supply for ITER all the current leads. There are three types of high temperature superconducting (HTS) current leads designed to carry 68 kA, 55 kA, or 10 kA to the ITER magnets. Different from the trial leads manufactured and tested in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), some critical technologies were qualified before the manufacturing of the ITER leads could commence, including: 1) welding, brazing, and soldering qualification; 2) fin type heat exchanger (HEX) with tight assembly tolerance; 3) low temperature superconducting (LTS) linker with cables soldered to the 5 K copper terminal; 4) instrumentation mock-up for the temperature and voltage test. In this paper, the outcome of these qualification activities will be reported together with the plans for the series production planning and development. [ABSTRACT FROM PUBLISHER]

• The ITER current lead insulation consist of interleaved pre-preg and polyimide tapes. • The acceptance high voltage test for current lead includes DC, Paschen and PD test. • ASIPP has completed the insulation qualification of ITER current lead. The ITER high temperature superconducting current lead is a critical component for the magnet system, which has the benefit of reduction in the heat load of the cryogenic system compare with the conventional lead. The current lead is located in the coil terminal box and dry box at the end of the ITER feeder system. As a warm to cold transition section, the current leads feed the huge current from the power supply system into the coils. At the outer surface of the current lead, one layer of a composite insulation is applied to isolate the high voltage potential of the current lead to the ground potential of the environment. The main body of the current lead is that of a long cylinder, but at the cold termination and the cooling inlet, the local geometry is much more irregular. So the insulation wrapping and curing technology of the current lead had to be developed to acquire its uniform mechanical and electrical performance. Now, the multi-stage autoclave curing technology has been qualified in ASIPP and the series ITER current leads are being manufactured based on the qualified procedure. In this paper, the latest insulation progresses for the ITER current lead are introduced, the high voltage testing results as part of the formal qualification are presented and discussed. [ABSTRACT FROM AUTHOR]

Heat conduction load test is a main and tough issue in the complex and huge general atom (GA) feeder system cold test. The GA feeder system is a superconducting system connects from the room temperature power supply system to the 4.5-K busbar of the magnet, supply 52 kA current and cryogenic provisions to the central solenoid magnet (CSM) coil, used for International Thermonuclear Experimental Reactor CSM magnet test at General Atomics Company, San Diego, CA, USA. This system have been developed and constructed at the Institute of Plasma Physics, Chinese Academy of Sciences, and completed the full cold test before transferred to the USA. In all the cold test items, the requirement of overall heat load 68 W is a big challenge, for its many leakage heat sources and cooling circuits: high temperature superconducting current lead, vacuum barrier, busbars, pipes, supports, etc. In order to get the steady results, many efforts were took to achieve the required value, such as reduce the inlet pressure of the 5-K loop as much as possible, minimize the flow rate of different 5-K cooling circuits, and at the same time, a dedicated program was developed to online calculate the density and the heat load while monitoring and analyzing the heat load. This paper will give a detail description of the 5-K overall heat load measurement. [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 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]

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]

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]

• A pair of 100 kA high temperature superconducting (HTS) current leads for CFETR is being designed at ASIPP. • The analysis of the maximum current carrying capacity of HTS module with self-field and external field has been performed. • The simulation of LOFA and burnout time for the 100 kA HTS current lead are also discussed. • The analysis results indicate that maximum current carrying capacity, LOFA and burnout time meet CFETR requirement. Based on the Agreement of Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Ministry of Science and Technology of China during the Thirteen Five-Years Plan (so called 13∙5 plan), a dedicated test facility of superconducting magnet for China Fusion Engineering Test Reactor (CFETR) will be constructed in Hefei, China. A pair of high temperature superconducting (HTS) current leads rated at 100 kA for the CFETR magnet test facility is being designed at ASIPP. Because magnetic energy stored in the toroidal field magnet is 134 Giga Joules which are 3 times higher than the ITER case, HTS current leads must work at nominal current without performance degradation during the HTS current lead quench. In order to verify the safety and reliability of HTS current lead, the analysis of static magnetic field is done by ANSYS iteration to predict the maximum current carrying capacity of HTS module. This paper also give the numerical calculation of loss of flow accident(LOFA) of HTS current lead and temperature margin analysis of HTS module. The analysis results indicate that the HTS module has a current carrying capacity of 165 kA without external field at 65 K and a current carrying capacity of 149 kA with external 30 mT vertical field at 65 K. The LOFA time is about 580 s and the burnout time is about 30 s, which is better than the requirement of the toroidal field coil fast discharge time of 24 s. [ABSTRACT FROM AUTHOR]