[en] Highlights: • The eddy current of CFETR vacuum vessel can be calculated by using a series of ideal current loops. • The shielding effect with different eddy current is studied by decomposing the exciting magnetic field as two orthogonal components. • The shielding effect can be determined from the rate of eddy current magnetic field to the external magnetic field. - Abstract: The operation of superconducting magnet for fusion device is under the complex magnetic field condition, which affect the stabilization of superconductor. The coil-case of TF coil can shield the magnetic field to some extent. The shielding effect is related to the eddy current of coil-case. The shielding effect with different eddy current is studied by decomposing the exciting magnetic field as two orthogonal components, respectively. The results indicate that the shielding effect of CFETR TF coil-case has obvious different with the different directional magnetic field, and it’s larger for tangential magnetic compared with that for normal field.

[en] Highlights: • A multi-scenario force-calculating simulator for Tokamak magnet system is developed using interaction matrix method. • The simulator is applied to EM analysis of CFETR and EAST magnet system. • The EM loads on CFETR magnet coils at different typical scenarios and the EM loads acting on magnet system of EAST as function of time for different shots are analyzed with the simulator. • Results indicate that the approach can be conveniently used for multi-scenario and real-time EM analysis of Tokamak magnet system. - Abstract: A technology for electromagnetic (EM) analysis of the current-carrying components in tokamaks has been proposed recently (Rozov, 2013; Rozov and Alekseev, 2015). According to this method, the EM loads can be obtained by a linear transform of given currents using the pre-computed interaction matrix. Based on this technology, a multi-scenario force-calculating simulator for Tokamak magnet system is developed using Fortran programming in this paper. And the simulator is applied to EM analysis of China Fusion Engineering Test Reactor (CFETR) and Experimental Advanced Superconducting Tokamak (EAST) magnet system. The pre-computed EM interaction matrices of CFETR and EAST magnet system are implanted into the simulator, then the EM loads on CFETR magnet coils at different typical scenarios are evaluated with the simulator, and the comparison of the results with ANSYS method results validates the efficiency and accuracy of the method. Using the simulator, the EM loads acting on magnet system of EAST as function of time for different shots are further analyzed, and results indicate that the approach can be conveniently used for the real-time EM analysis of Tokamak magnet system.

[en] Highlights: • The eddy current distribution and variation of CFETR vacuum vessel during plasma disruption have been calculated. • Effective electrical parameters can be derived from the eddy current characters. • The method for eddy current and effective electrical parameters is suit for the complex shell with arbitrary shape. - Abstract: The electrical parameters of CFETR (China Fusion Engineering Test Reactor) vacuum vessel are very important to the design of control system and power supply system. Effective electrical parameters are relevant to the dynamic of eddy current. For complex structure, the distribution of eddy current can’t be obtained by analytical form. A method is presented to solve the eddy current of the vacuum vessel in this paper. The effective electrical parameters can be got from the eddy current distribution and variation. The time constant of the CFETR vacuum vessel is derived from the decay characteristics of the eddy current. And the effective resistance and inductance can be derived from the viewpoint of energy for a certain distribution of eddy current.

[en] Highlights: • A new technology–interaction matrix method is applied to assess EM loads of EAST magnet system. • The interaction matrices of EAST magnet system are obtained. • The application validated the efficiency and accuracy of the method. • Results indicate that the approach can be conveniently used for multi-scenario EM loads assessment for EAST current-carrying components. - Abstract: An approach for assessing the electromagnetic (EM) loads of the main current-carrying components in tokamaks has been proposed recently [1,2]. It is mainly based on the interaction matrix and the method is general. This paper explores on the application of the new technology to EAST magnet system. Firstly, the interaction matrices of EAST magnet composed of bilateral interaction forces between separate components at unit current are calculated, then the EM loads are obtained by a linear transform of given currents using the interaction matrix. The application validated the efficiency and accuracy of the method, which is useful for the systematic assessment of Tokamak EM forces. Results indicate that the approach can be conveniently used for multi-scenario EM assessments and parametric studies of the EM loads for EAST current-carrying components, and a specialized force-calculating module for real-time simulating will be developed in the future.

[en] Correction coils (CC) are the important components of International Thermonuclear Experimental Reactor (ITER) superconducting systems. The coil body is composed of many types of material and periodic micro-structure. It is important to predict its equivalent material properties. A method commonly used is finite element method (FEM) based on homogenization theory, whose process is complicated. A new FEM based on Generalized Hook's Law was proposed. Compared two methods with each other,it is found that the latter is easier and more precise. The results provide essential parameters for structure analysis and thermal analysis for CC. (authors)

[en] Integrating engineering software is meaningful but challenging for a system code of a fusion device. This issue is seldom considered by system codes currently. Therefore, to discuss the issue, the Integrated Design System of TF Coil (IDS-TFC) has been worked out, which consists of physical calculation, CAD, and Finite Element Analysis (FEA). Furthermore, an Integrated and Automatically Optimized Method (IAOM) has been created to address the integration and interfaces. The method utilizes a geometry parameter to connect each design submodule and achieve automatic optimization. Double-objectives optimization has been realized, confirming it is feasible to integrate and optimize engineering design and physical calculation. Moreover, IDS-TFC can also serve as a useful reference of integrated design processing for subsequent fusion design. (paper)

[en] The Correction Coils (CC) in ITER consist of 18 individual coils arranged in 3 groups around the toroidal field coils(TF), those are 6 on top(TCC), 6 on side(SCC) and 6 on bottom(BCC) by out of the TF. The correction coils are placed and sized can correct the most significant error on poloidal (PF), TF and feeders that come from manufacturing and assembly tolerances. These opposite pairs of coils are connected in 9 independent circuits. The coils have been designed using a cable-in-conduit NbTi conductor (CICC) with a 10 kA operating current. They are pancake wound with a vacuum pressure impregnated (VPI) with glass-kapton-epoxy insulating system. The winding pack is contained in a case which is mounted onto the TF coil cases as support. The ITER Organization (IO) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly preparing the definition of the technical specifications and the upcoming qualification program for the Correction Coils. The development of major items such as terminal joints, casing manufacture, and vacuum impregnation system, is an essential phase before the series production which will take place at the premises of the supplier. This paper was shown the status on key technologies for fabricating of CC coils. (author)

[en] The Correction Coils System (CCs or CC) in ITER consists of eighteen individual multi-turn coils arranged in three groups (Top, Side and Bottom) around the TF coils. It was used to compensate field errors arising from misalignment of the coils and winding deviations from the nominal shape as a result of fabrication tolerances, joints, leads and assembly tolerances. The six side coils are also used for feedback control of plasma resistive wall mode (RWM) disturbances. The top and bottom coils is located between TF cases and PF6 coil and consists of six saddle type coils covers a 60deg sector and spans three TF coils in the toroidal direction at upper and lower of the TF coil. The side coils is located between TF cases and both of PF3 coil, PF4 coil and consists of six quadrilateral shape type coils arranged around the toroidal circumference with a 40deg sector and each coil spans only two TF coils at aside of the TF coil. All Corrections Coils (lower, side and upper) are then attached to the TF coil cases. The electromagnetic field and forces acting on the CC winding are calculated with the CC current and ITER presented design scenario (burn time), which includes TF in operating, PF system discharge, plasma initiation, current ramp-up and heating to a driven burn. The CCs experience electromagnetic forces due to their self-field and the field of the TF coils, PF coils and CS coils. During calculating, the geometrical dimensions of all coils are originally defined in the reference of ITER Design Description Document DDD 11 Magnet Section 1. It was an engineering description (ITER D_--22HV5L v2.2) in 2006 version. (author)

[en] Highlights: 1Adding tungsten armor increases the eddy current density in the first wall. 2Adding tungsten armor increases EM force and moment on the blanket. 3The thickness of the tungsten armor increases the eddy current density. 4Split size increases the eddy current density, temperature and stress. 5Larger gap can decrease the eddy current density. Tungsten armors are arranged on the plasma side of the blankets to protect the first wall in a fusion device, such as China Fusion Engineering Test Reactor (CFETR) and European Demonstration Power Plant (EU DEMO). Tungsten armors will produce a huge eddy current and Electromagnetic (EM) load under plasma disruption due to their high conductivity and strong magnetic field environment, which will cause the blankets to be subjected to a substantial eddy current and thus heat load, thermal stress and mechanical stress, that potential could damage blankets. To evaluate the effect of the tungsten armor on the electromagnetic characteristics of blanket, electromagnetic calculation method verification is carried out. And the electromagnetic finite element model of blanket is established by using ANSYS, and the eddy current distribution in the first wall is calculated. Effects of the tungsten armor, split size and gap size of the tungsten armor on the eddy current of blanket first wall and effect of the tungsten armor on the EM force of blanket are researched. And the influence of different tungsten armor design on the thermal stress distribution of blanket first wall is also studied. The study results will provide an important reference for the blanket design of the fusion device.

[en] ITER edge localized mode (ELM) coils are important components of the in-vessel coils (IVCs) and they are designed for mitigating or suppressing ELMs. The coils located on the vacuum vessel (VV) and behind the blanket are subjected to high temperature due to the nuclear heat from the plasma, the Ohmic heat induced by the working current and the thermal radiation from the environment. The water serves as coolant to remove the heat deposited into the coils. Based on the results of nuclear analysis, the thermal-hydraulic analysis is performed for the preliminary design of upper ELM coils using a rapid evaluation method based on 1D treatment. The thermal-hydraulic design and operating parameters including the water flow velocity are optimized. It is found that the rapid evaluation method based on 1D treatment is feasible and reliable. According to the rapid analysis method, the thermal hydraulic parameters of two water flow schemes are computed and proved similar to each other, providing an effective basis for the coil design. Finally, considering jointly the pressure drop requirement and the cooling capacity, the flow velocity is optimized to 5 m/s. (fusion engineering)