Fusion Engineering and Design最新文献

This study describes the design and development of double-walled bellows used in ITER's vacuum ultra-violet (VUV) spectrometer systems. These bellows were designed to accommodate port plug displacement resulting from the thermal expansion of the ITER vacuum vessel and other interface loads. The three VUV spectrometer systems under development—the VUV core survey spectrometer, divertor-VUV spectrometer, and VUV-edge spectrometer—are intended to measure the VUV spectrum to monitor impurity species in the plasma. While the ex-vessel components of ITER VUV spectrometers in the interspace and port cell areas, are fixed to the tokamak building, the closure plate of the port plugs, to which the sight pipes are attached, is expected to move by several centimeters both vertically and radially owing to the vacuum vessel's thermal expansion. As a result, the structural integrity of the system requires a flexible structure that can compensate for displacement caused by thermal expansion. To address this need, two types of double-walled bellows were developed, gimbal bellows for lateral displacement and axial bellows for axial displacement. These were developed in accordance with the ASME SEC.VIII code and EJMA standard to meet ITER's safety and vacuum requirements. A double wall is mandated for vacuum bellows due to their vulnerability in terms of vacuum safety. The structural integrity and functionality of the designed bellows were confirmed through functional tests on the manufactured prototype.

Effective project management methods, tools and working practices shall be applied to facilitate the communication and collaboration among the different institutions involved in the Divertor Tokamak Test (DTT) project. This paper deals with the definition of the configuration management platform architecture enabling technical integration of the DTT system. The first step consists in the identification of main requirements the platform should satisfy, considering the multidisciplinary domains and the geographically dispersed working teams characterizing the nuclear fusion sector and the maturity level of the specific project. Main characteristics of the most advanced Product Lifecycle Management (PLM) tools are identified, their limits and benefits are evaluated, and the suitable PLM platform is selected, satisfying the DTT project needs. The definition of the architecture of the configuration management platform for the DTT project aims at implementing the DTT assembly model in a unique environment able to exchange models and data even developed outside the platform ensuring the congruence of the design, the traceability of design changes and the adoption of a proper Systems Engineering approach.

A field-programmable gate array (FPGA)-based trigger system (FTS) is designed and developed to replace the existing microcontroller-based system for the automation of metallic tokamak (MT-I). MT-I consists of three main systems: MT-I vacuum vessel and gas filling system, a pulsed power supply system, an electronic system based on diagnostics and data acquisition (DAQ) systems. The power supply system provides pulsed power input to the central solenoid (CS), poloidal field coils (PF), toroidal field coils (TF), and microwave systems using thyristors and insulated-gate bipolar transistors (IGBTs). The (DAQ) system acquires experimental data from MT-I diagnostics using national instruments (NI) DAQ cards. A concurrent processing system is required to incorporate time delay in the triggering process of central solenoid (CS), poloidal field coils (PF), toroidal field coils (TF), microwave source and DAQ systems. Therefore, an FTS is designed and developed to complement the processing capability, unlike a microcontroller. The FTS has twenty concurrent triggering channels, adjustable from time reference zero, and control from a simple graphical user interface (GUI) designed on LabVIEW. For current buffering and optical isolation against high voltages in the MT-I power supply system, a peripheral electronic system (PES) and field trigger modules (FTM) have been developed as part of FTS. The designed and developed FTS was tested successfully on MT-I.

HL-3 is a tokamak device constructed in 2020 and it has achieved remarkable results in a short period of time. Baking is an important way to obtain a high vacuum required for plasma discharge in tokamak devices. A nitrogen gas closed-circuit reheat baking system is utilized to concurrently bake the vacuum vessel (VV) and the lower divertor of HL-3. So far, three rounds of baking experiments have been carried out, with vacuum pressure up to 1.8 × 10^–6 Pa and the results from the mass spectrometer indicated that the amount of compounds were greatly reduced in the residual gas. Besides the design description and experimental results of HL-3 baking system, this paper also introduces a simulation-based solution for addressing the issue of the lower divertor temperature being significantly lower than that of the VV after baking to the high temperature stage.

As a plasma-facing material (PFM), W–ZrC is a promising candidate for the divertor in future fusion reactors. To verify the feasibility of W–ZrC, a high heat flux (HHF) test was conducted on a comprehensive experimental platform. A rare internal ablation failure mode was observed during the experiment. Considering the internal ablation phenomenon during the test, the measured data of the test platform were analyzed to predict an ablation time of 12 s. Based on the shape and size of the ablation pit, the cause and mechanism of ablation were determined using thermal simulation analysis. It can be concluded that a sudden and significant increase in heat flux combined with reduced water flow can cause thermal ablation to occur within the heat sink. Furthermore, dimensions of the ablation pit are related to the area in which the heat flux is imposed. In sum, internal ablation can have serious consequences for the divertor.

One of the significant obstacles left if the development of a fusion power plant is the development of Plasma Facing Components (PFCs) that can withstand the large heat and particle flux incident on the first wall and divertor region. As solid PFCs struggle with microstructure growth, sputtering and melting, liquid metals have been a popular potential replacement. Liquid metal's use as a PFC has been increasing due to its ability to self-repair damage as well as pump lost fuel and waste particles to create a low recycling edge. Currently, tin, lithium and lithium eutectics are the commonly considered liquid metals for use as a fusion PFC. It is therefore important to research the Plasma Material Interaction (PMI) between a hydrogen plasma and the liquid metal PFC candidates. This work, conducted at the Center for Plasma Material Interaction (CPMI), investigated how a molten tin surface reacts when exposed to a hydrogen plasma, building off observations by ASML of particles being ejected from a molten tin surface in the presence of a hydrogen plasma. With the ejection of tin affecting ASML's systems as well, since ejected tin can travel around their systems and cause contamination of equipment or wafers that can be destructive to the lithography process. So, for this work, molten tin was exposed to a hydrogen plasma, of varying electron densities and temperatures, and any ejected particles were collected on a witness plate to determine the particle sizes, angular distribution and mass flux. This work found the ejected tin particles range in size from 10′s of$nm$s to 10′s of microns and that the mass flux of tin from the molten surface is in the 10′s of $\frac{mg}{m^2*s}$, increasing with increased atomic hydrogen flux to the molten tin surface. Showing macroscopic amounts of molten tin are ejected in the presence of a hydrogen plasma, therefore showing tin may not be a suitable fusion PFC material.

The solid breeder blanket is a critical component in fusion reactor, where helium purge gas flows through the solid breeder pebble bed to carry out the tritium generated during the fusion process. The flow pressure drop of helium purge gas within the breeder pebble bed is a significant parameter affecting the design of the tritium extraction system. Previous studies have indicated that the helium flow in the breeder pebble bed conforms to the theory of porous media flow. However, due to potential pebble breakage during the plasma operation, the pressure drop characteristics of the helium flow in breeder pebble bed may change as the void structure changes. The objective of this study is to measure the variation in flow pressure drop of the breeder pebble bed under different pebble crushing conditions. The flow pressure drops of intact beds (Alumina, diameter 1-1.2 mm) and four groups with different pebble breakage rates (3%, 5%, 7%, 9%) are measured using the pebble bed pressure drop test facility. The following results are obtained through experimental research: (1) The Ergun equation, Foumeny equation, and Reichelt equation can all reasonably match the experimental results of intact pebble beds; (2) The pressure drop across the pebble bed increases with the increase in pebble breakage rate, reaching approximately 1.6 times that of the intact bed at a 9% breakage rate; (3) A correlation for predicting the pressure drop of the broken pebble bed is established by introducing the pebble breakage rate ($\eta$) into the Ergun equation, which can be used to determine the pressure drop variation within a conservative range of breakage rates.

The Liquid Lead lithium Magneto Hydro Dynamics (LLMHD) experimental facility has been constructed at Institute for Plasma Research (IPR), Gujarat, India to perform various R & D MHD experiments associated with the flow of electrically conducting liquid metal under strong transverse magnetic field. The electromagnet having C-shaped soft iron core has been designed and developed, to provide a uniform magnetic field of up to 1.4T within its polar volume 1000 mm (H) ×400 mm (W) ×370 mm (L). The magnetic field lines are aligned along the length (L). A relatively large polar volume inside the electromagnet to place the test mock up for MHD experiments is its particularity. It enables the study of MHD flows with complex flow geometries and having longer flow length perpendicular to the magnetic field. We have started running the LLMHD loop and the first MHD experiments with Pb-Li have been performed so far at 320 °C in a test mock-up of a basic circular flow geometry having two 90° bends. So far till now, the isothermal MHD experiments have been conducted in the presence of a uniform transverse magnetic field of 0.62T (Ha ∼ 322) and 1.06T (Ha∼551) for the ranges of Reynolds number 20,000–50,000. During the MHD experiments, flow rates, temperature, pressure, and induced wall electric potential have been recorded. The MHD effects on the pressure drop and flow rate has been noticed. The 3D MHD numerical simulation has also been performed, using add on MHD module of ANSYS FLUENT. Both simulation and experimental results of the induced wall electric potential have been compared.