Fusion Engineering and Design最新文献

High-power switches, as the important active part of Fast Discharge Unit (FDU) in Tokamak, are designed to interrupt the circuit current and transfer the quench energy for protecting expensive superconducting magnets. Hence, their reliability has always been highly underlined. In this paper, the reliability and test number of FDU switches are discussed, mainly including three issues. First, the FDU configuration characteristics of different types of superconducting magnets are introduced, and explanations and calculations on reliability are provided. Second, the reliability interval estimation after a certain test number is discussed by using the binomial distribution theory. Three, the compliance test based on the claimed reliability is studied, and two types of test plans are discussed and compared to derive the test number. The discussion in this paper focuses on the FDU switches and provides valuable guidance for their reliability test and assessment.

Engineering design of the JA-DEMO divertor concept, i.e. double-coolant circuits of 200°C coolant (5MPa) for high heat load targets (CuCrZr-pipe) and 290°C coolant (15MPa) for high neutron load Plasma Facing Units (F82H-pipe) and cassette body (CB), has been developed. Computational fluid dynamics (CFD) calculation of the coolant distributions to targets, baffles, reflectors, dome and to CB was performed in order to determine the feasibility of the key concepts. (i) All PFU support designs for the coolant distribution to PFUs were optimized to reduce variation of the flow velocity (V_cool) at the inlet of PFUs less than 5 – 6 %. Parallel coolant circuit design for the dome and reflectors was also developed, and mass flows were adjusted by orifices and the main mass flow rate. (ii) A new cooling concept with two layers of puddles and fins at the side routes was proposed for CB. The design provided V_cool = 0.55 – 1.27 m⋅s⁻¹ and average V_cool = 1.04 m⋅s⁻¹ with the fin transparent ratio of 0.1. It was enough to exhaust the nuclear heat, and the design issues were identified under the coolant condition. (iii) Heat exhaust on fish-scale surface target and stress-strain of the CuCrZr pipe were investigated under assuming a DEMO divertor condition. Steady-state heat load at wet region by plasma (q_t^wet) was restricted below 13.5 MWm⁻² to avoid W-recrystallization. In the stress-strain cycle of higher q_t^wet ∼15.3 MWm⁻², maximum tensile σ (120 – 200 MPa) and total change in strain during the heat load cycle (Δε ∼0.25 %) were relatively small. These evaluations suggested that such high heat load cycle may not be a critical lifetime issue, but reduction in T_cool is preferable to handle ITER-like slow transients such as 15 – 20 MWm⁻².

The paper presents and discusses the results of the research on the preparation of the HF plasma heating system of the KTM tokamak for the start-up. The calculations have been performed to determine the parameters at which the most effective absorption of the HF radiation by the plasma of the KTM tokamak was expected, the results of measurements of the frequency response of the antenna-feeder device of the HF heating system generator of the KTM tokamak plasma have been presented. The result of the carried out work have shown that more efficient plasma heating on a KTM tokamak can be obtained at a frequency of 14 MHz.

This paper introduces the safety and protection measures implemented in the High Heat Flux Test Facility (HHFTF). The interlock system has been enhanced and integrated into HHFTF to ensure the protection of the facility and the safety of personnel, thereby mitigating risks during operations. The primary purpose of these interlocks is to secure the entire facility under well-defined conditions to prevent accidents.

The safety and interlock protection system constitute integral components of HHFTF's overall control system, providing the capability to manage various subsystems, many of these systems are operated reliably close to their performance limits in order to achieve the desired goals. The overall safety, interlock protection and the control system encompasses both hardware components and software, employing Programmable Logic Controllers (PLCs) and Field Programmable Gate Arrays (FPGAs), along with wired logic based on relays and special logic cards. Three different types of architecture have been developed: (1) Slow Architecture based on PLCs, for functions where response time of longer than 20 ms is adequate; (2) Fast Architecture based on FPGAs, for functions requiring fast response time beyond the capabilities of the PLC; and (3) Hardwired Architecture for critical functions.

The paper showcases the successful testing and outcomes of an enhanced interlock protection system, encompassing both software and critical hardwired interlocks, within the HHFTF. Key parameters monitored include the maximum allowable job threshold temperature, flow rates (for coolant loss detection), and chamber pressure. Activation times of interlocks were observed within a range from 100 microseconds to 116 milliseconds.

In EAST(Experimental Advanced Superconducting Tokamak), the heat produced by ECRH (Electron Cyclotron Resonance Heating) and ICRF (Ion Cyclotron range of Frequencies Heating) should be respectively removed through independent cooling water system. Their capacities are typically proposed based on the full power of the ECRH and ICRF. Indeed, the operational duration of the wave system is limited, which necessitates a larger cooling water capacity redundancy, occupies a significant amount of space, and requires a substantial budget. In this paper, the cooling water system for ECRH and ICRF of CRAFT (Comprehensive Research Facility for Fusion Technology) testing bench is designed according to its requirements.Initially, the optimization of the cooling water process is carried out in accordance with the specified parameters. Thereafter, the entire process is simulated employing the AFT Fathom code.Finally, a multiuser integrated cooling water system is designed, which is different from the traditional majority one-to-one mode. And the results show that all the parameters can meet the system requirements, meanwhile the redundancy of the cooling water capacity can be greatly reduced by flow regulation.The cooling water integral structure will provide a new idea for the design of fusion reactor cooling water systems.

KTX, a reversed field pinch (RFP) device, is equipped with a robust Magnetohydrodynamic (MHD) magnetic field feedback control system, which comprises two distinct loops: an inner loop used for current control to generate a magnetic field, and an outer loop dedicated to magnetic field control; these loops operate in a cascading manner. The Proportional–Integral–Derivative (PID) control methodology is employed. In the pursuit of achieving superior performance, mathematical analysis is conducted on both control loops. This rigorous examination culminates in the design of an Internal Model Control (IMC) PID controller. Simulations are employed and further verified through experimental investigations. A magnetic field with fast and accurate response time has been successfully achieved.

A beam driven plasma neutraliser (BDPN) has been proposed [1] as a relatively simple way to increase the efficiency of neutral beam injectors based on the acceleration and neutralisation of H^- or D^-. Initial calculations showed that sufficient levels of ionisation could be achieved with H₂ or D₂ as the initial gas target in the neutraliser if a plasma confinement time of the plasma in the neutraliser is >0.5 ms could be achieved. However, later calculations [2,3] show that the afore-mentioned results were very optimistic as the calculations did not take into account the recombination of molecular ions with electrons. Hence, it has been suggested [4] to use an argon (Ar) plasma as the molecular ion content in such a plasma will be negligible. The use of an Ar plasma neutraliser has been suggested previously [[5], [6], [7]], but few details have been given of the obtainable neutralisation efficiency, the necessary line density for that neutralisation efficiency, or of the calculations and cross sections used to deduce those quantities. This paper briefly discusses why Ar could be a good choice for a plasma neutraliser, then the reactions occurring between the H^-/D^- beam and the particles in the Ar plasma, and the available cross section data for those reactions. Subsequently, the differential equations for the change of the species in the beam as it traverses an Ar plasma are deduced and solved, and the solutions used to calculate the achievable neutralisation efficiency as a function of the line density in the neutraliser etc. Results are given of the neutralisation efficiency as a function of the degree of ionisation in the neutraliser, the required line density, and the species changes in the beam for accelerated H^- beams with 100, 500 and 870 keV energies.

Hydrogen isotope permeation through the structural walls of a fusion reactor poses concern for both fuel loss and safety. A primary focus in the International Thermonuclear Experimental Reactor (ITER) is the reduction of hydrogen isotopes, particularly tritium permeation through structural materials. In this work, an experimental setup has been developed to study the permeation of deuterium through thin samples such as bare and erbia (Er₂O₃) coated 100 µm thick stainless steel (SS 316 L). Permeation results through a bare SS 316 L sample yielded deuterium diffusivity and permeability of SS 316 L. Erbia coating on SS 316 L was developed using dip coating technique and its performance was tested as a permeation reduction barrier (PRB). A permeation reduction factor of ∼87 and 359 have been achieved with coating thickness ∼ 100 and ∼200 nm respectively. No measureable permeation flux was obtained for coating thickness of 492 nm. These results indicate the effectiveness of erbia coating in reducing deuterium permeation through SS 316 L.

In this work, we present a general control architecture for robotic systems dedicated to the remote handling of in-vessel components in fusion machines. This control architecture will be tested for the inspection and maintenance of the first wall components of the RFX-mod2 experiment, in the scope of the New Equipment For the Experimental Research and Technological Advancement for the Rfx Infrastructure (NEFERTARI) project. The architecture is split into low-level and high-level layers. The former is used for the low-level control of the robotics systems and to manage safety; the latter is used for the high-level planning and implements several sub-modules, such as a Virtual Reality (VR) environment for the visualisation of the robot’s digital twin. Moreover, an Application Programming Interface (API) is intended to connect the two layers, and the communication between modules and layers is provided by the ROS2 framework. A typical usage of such a framework involves a human operator who is teleoperating the real robot, data from motor encoders are used as inputs for the dynamic model module. This module is used to compute the forward dynamics in real-time, providing an accurate simulation of the robot (digital twin) in the virtual environment.