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A China Fusion Engineering Testing Reactor (CFETR) will be built to test and verify the feasibility of engineering and technology in practice for the future fusion reactor. Long pulse and steady-state operation will be demonstrated with duty cycle time more than 30%–50%. Because of the beta and gamma activation of the component bulk and surface dust (beryllium, carbon, and tungsten), special remote handling (RH) techniques will be required during machine maintenance periods. The divertor cassettes’ remote replacement is one of the key maintenance operations for the CFETR to meet the requirements of duty cycle time. RH maintenance strategies will have a significant impact on the layout of the machine and design of components. In this article, an overview of the CFETR divertor components’ remote maintenance strategy and the description of the maintenance process of the divertor cassettes were introduced. The engineering design features of the divertor RH systems (DHRSs) were described to meet the requirements such as high efficiency and reliability. Preliminary assessment of the divertor RH system is done in order to successfully and efficiently carry out RH maintenance tasks. [ABSTRACT FROM AUTHOR]

In order to meet the heating requirement of EAST tokamak, ion cyclotron range of frequency (ICRF) heating is utilized as one of the main auxiliary heating methods, which plays an important role in coupling radio frequency (RF) power into the plasma. As a member of plasma facing components, ICRF antenna suffers the plasma radiation as well which is one of the causes of the thermal load on the antenna. Plasma radiation and RF thermal load during the operation of ICRF and thermal load caused by HALO current during vertical displacement event will contribute together to a huge thermal load on the surface of the antenna and consequently the structural stability and reliability of ICRF antenna will be affected. Therefore, cooling channels is needed to remove heat on the ICRF antenna. In this paper, cooling channels for straps of ICRF antenna are designed based on the thermal load on the antenna, and thermal-structural analysis of straps with cooling channels is also carried out by the finite-element method. [ABSTRACT FROM AUTHOR]

The new tokamak generation will be characterized by the necessity of full remote maintenance for most of the critical components. Because no human intervention can be envisaged, the remote systems will have to prove high reliability and rescue capabilities. The availability of the fusion facility will have to be maximized, as a consequence, the efficiency of the maintenance equipment will become a key factor. To fulfill these constraints, the remote maintenance principles will need to be simplified as much as possible and be taken into account at the early stages of the tokamak design. This is an iterative process between Tokamak and Maintenance designs in order to reach the best possible tradeoff. Such an exercise of this iterative work was done on the China Fusion Engineering Test Reactor (CFETR) preliminary design. Main targets were established in the view of optimization of the remote maintenance, then an assessment of the first maintenance scenario, envisaged for the in-vessel components, was done and alternative solutions were proposed. This process was repeated during meetings with the tokamak subsystem owners. This paper will present the results of a first phase of optimization that can represent the current state of the remote maintenance design for CFETR. The work has provided guidelines for main tokamak subsystems such as the magnet configuration, the blanket and divertor modularity, the neutron shielding and cryostat arrangement, and the transfer and hot cell configuration. It also produced some innovative solutions for the overall maintenance scheme like the implementation of a hot cell at the top of the tokamak hall avoiding cask transfers. On this basis a CFETR maintenance scenario is proposed. [ABSTRACT FROM AUTHOR]

The latest superconducting magnets (SM) for fusion are mostly force-cooled, mainly because it allows reliable electrical insulation of the coils using vacuum pressure impregnation (VPI). SM of this type have many leads, feeders and coolant tubes, located in cryostat vacuum, which must sustain high voltages, induced on them by fast current changes. However vacuum loss can spoil their insulation. A few such cases occurred during the T-15 tokamak coils testing, initially having bare leads relying upon vacuum. But its loss generated a coil quench, a protecting current dump at high voltage, followed by breakdown and arc. Even leads insulation by Teflon and fiberglass tape wrap proved to be insufficient. Nevertheless, similar tape wrap insulation of leads and feeders (ILF) was used in EAST, KSTAR, SST-1 and W-7X. So far, seven breakdowns occurred during their coil tests at operating voltage \sim<3 kV. Breakdowns never initiated in the coils, but always on their leads, feeders and sensor lines, indicating that their insulation made by tape wrap were too weak. Instead of ILF improvement some projects undertake Paschen tests. These are planned as the baseline for ITER too. But these tests are valid for the coil with open insulated surface, but are not appropriate for the final tests, when insulation should not be exposed to vacuum. Up to now ILF final tests have been done in all devices at 10–21 kV, but only in good vacuum in spite of the fact that such tests could not guarantee safe operation in case of vacuum loss. We propose to increase ILF strength to the same level, as in the coils, using vacuum-tight grounded stainless steel casings filled up by VPI over magnet leads. This will provide reliable and easily testable solid insulation. Besides, casings would exclude He leaks, providing the second vacuum tight barrier over the ILF. Thus it would increase the magnet reliability and would make it possible to avoid the needs of all single coils test. [ABSTRACT FROM PUBLISHER]