Vacuum Vessel Research Articles

Power handling in a highly-radiative negative triangularity pilot plant

Abstract This work explores detailed power handling solutions for a class of high-field, highly-radiative negative triangularity (NT) reactors based around the MANTA concept [Rutherford et al. (2024)]. The divertor design is kept as simple as possible, opting for a standard divertor with standard leg length. FreeGS is used to create an equilibrium for the boundary region, prioritizing a short outer leg length of only ∼50 cm (∼40% of the minor radius). The UEDGE code package is used for the boundary plasma solution, to track plasma temperatures and fluxes to the divertor targets. It is found that for PSOL = 25 MW and nsep = 0.96 × 1020 m−3, conditions consistent with initial core transport modeling, little additional power mitigation is necessary. For a fixed impurity fraction of just 0.13\% Ne in the plasma, the peak heat flux density at the more heavily loaded outer targets falls to 7.8 MW/m2, while the electron temperature, Te, remains just under 5 eV. Scans around the parameter space reveal that even at densities lower than in the primary operating scenario, PSOL can be increased up to 50 MW, so long as a slightly higher fraction of extrinsic radiator is used. With less than 1% neon (Ne) impurity content, the divertor still experiences less than 10 MW/m2 at the outer target. Design of the plasma-facing components includes a close-fitting vacuum vessel with a tungsten inner surface as well as FLiBe-carrying cooling channels fashioned into the VV wall directly behind the divertor targets. For the seeded heat flux profile, Ansys Fluent heat transfer simulations estimate that the outer target temperature remains at just below 1550◦C. Initial scoping of advanced divertor designs shows that for an X-divertor, detachment of the outer target becomes much simpler, and plasma fluxes to the targets drop considerably with only 0.01% Ne content.

Research on the inverse kinematics of heavy-load manipulator and end-effector for fusion reactor

The environment in Tokamak fusion reactor is hostile to remote handling (RH) equipment that maintains in-vessel components. Many maintenance tasks include transferring components, pipe cutting, and welding in the vacuum vessel (VV). To fulfill these maintenance works, a 7- degree-of-freedom (DOF) heavy load manipulator and a 3-DOF end-effector are designed and connected together to form a 10-DOF redundant robot. The inverse kinematics for such a high redundancy robot, considering obstacle avoidance in the complex space of VV, is a challenge. To solve this problem, the Jacobian method is adopted, and a method specially designed for obstacle avoidance of redundant robots working in the VV is integrated to simplify algorithm calculation and improve inverse solution efficiency. The algorithm is applied to solve the inverse kinematics of the redundant robot for the maintenance of the divertor plasma facing components (PFC) in VV. The results show that the inverse kinematics of the redundant robot can be solved quickly with high positioning accuracy. The method can also be used for other fusion reactor maintenance robots with some adjustments. The maintenance efficiency of redundant robot can be improved with the method which can be valuable for the whole remote handling process.

Design and test of a module of a breathable Electrostatic Shield for the MITICA 1 MV negative ion Beam Source

The electrical insulation of the MITICA Beam Source at 1 MV is a challenging issue, which has not been fully addressed so far on the basis of experimental results and of theoretical models available in literature. Being MITICA the full-size prototype of the Heating Neutral Beam Injector for the ITER fusion experiment, its electrical insulation is constituted just by vacuum gaps and alumina insulators, since other insulating materials such as SF6 gas or fibreglass-reinforced plastic (FRP) would be quickly degraded by the expected neutron flux produced by fusion reaction. Extrapolations based on HV tests on reduced-scale models have recently indicated the risk of electrical breakdowns in the vacuum gap between electrodes nominally operating at -1 MV and the vacuum vessel (at ground potential). The risk of electrical breakdown can be mitigated by introducing an intermediate Electrostatic Shield (ES), which essentially is an equipotential (metallic) enclosure surrounding the HV electrode, so as to divide the vacuum gap in two independent insulating gaps of 400 kV and 600 kV respectively. However, for optimal negative ion production, the ion source shall operate in H2 or D2 at a pressure of ∼ 0.3 Pa and unavoidably produces a flow of gas leaking out in the surrounding vacuum. Thus, the presence of an intermediate shield can substantially increase the background gas pressure in the vacuum gaps, and, due to the large gap length (0.6 m), exacerbate the risk of breakdown when the pressure approaches the conditions of Paschen-type discharges. In addition to this, RF-induced breakdowns were found on the rear side of the ion source during the operation of the prototype source SPIDER, which were somewhat correlated to a relatively high hydrogen pressure in that area. For these reasons, a structure capable of constituting a full equipotential barrier all around the BS and, at the same time, having sufficient gas conductivity (breathability) to allow efficient pumping of background gas, has been designed. In the first part of the paper, the requirements and design optimization of a breathable module of the intermediate ES are described. Then, an experimental campaign for the validation of the electrode implementation the test configurations and the experimental procedure is discussed.

Development of a thermal diagnostic system for the SPARC tokamak.

The SPARC tokamak will have scrape-off layer parallel heat fluxes on the order of GW/m2. Managing power exhaust of this magnitude will be mandatory for a reactor-scale device. To enable this mission, a thermal diagnostic suite will be deployed to measure the in-vessel structural temperatures to ensure they do not exceed their design limits and to determine the spatial distribution and magnitude of energy deposited onto the first wall. Thermocouples and fiber Bragg gratings have been selected for their environmental compatibility and proven useful on other fusion devices. High-density thermocouple arrays in the divertor will have two spring-loaded thermocouples per divertor target tile, which are being used as calorimeters, and will look to resolve the temperature distribution within the tile due to a swept or static strike point. All systems will need to survive the vacuum vessel bake, set at a minimum plasma facing surface temperature of 350 °C, which presents a particularly challenging environment for the fiber-based subsystem. Along with this temperature design requirement, all the materials in the primary vacuum need to be ultra-high vacuum compatible, able to handle the expected neutron and gamma radiation, as well as tritium exposure, all of which restrict material options. Finally, due to the expected activated environment in SPARC, there will be little chance to replace defective sensors, so system resilience is ensured through toroidal redundancy, probe material selection, and mitigating the impact of common-mode failures. Initial testing of sensors show that intershot structural measurements are sufficiently captured with the raw output, but intrashot measurements of the plasma facing material requires model-based interpretive tools.

Application of a non-intrusive reduced order modeling approach to magnetohydrodynamics

Magnetohydrodynamics (MHD) investigates the intricate relationship between electromagnetism and fluid dynamics, offering a complete insight into the behavior of conducting fluids under the influence of magnetic fields. This theory plays a pivotal role in the framework of magnetic confinement fusion, where it can be applied to describe both thermonuclear plasmas confined inside the vacuum vessel and operating fluids, such as liquid metals and molten salts, flowing within the blanket of future tokamaks. Currently, the state-of-the-art numerical modeling of MHD scenarios employs a multi-physics framework to examine the interplay between magnetic fields and thermal hydraulics; however, due to the complexity of the involved physics, detailed models are required, resulting in a significant computational burden. In this regard, reduced order modeling (ROM) techniques may represent a promising solution, as they enable approximating complex systems with lower-dimensional models. Indeed, ROM methodologies can significantly reduce the required computational time while maintaining accuracy in capturing the convoluted physics involved in fusion reactors, especially in the contexts of sensitivity analysis, uncertainty quantification, and control. Despite their potential, ROM methods are relatively under-explored within the MHD framework; this study applies ROM techniques to MHD scenarios, focusing on their capabilities and possible limitations. To this aim, the backward-facing step, which is well suited for exploring the effects of different magnetic fields on turbulent dynamics, is adopted as case study. In particular, this work evaluates the potentialities of the ROM approach in enhancing computational efficiency within the MHD domain. Each of the methods evaluated was effective in precisely reconstructing flow dynamics at any given time and across the full range of magnetic field values tested while significantly reducing computational costs compared to full-order simulations. Practically, this study demonstrates the feasibility to create simplified models that accurately represent the magnetohydrodynamic flows of fluids within the blanket.

Overview of the early campaign diagnostics for the SPARC tokamak (invited).

The SPARC tokamak is a high-field, Bt0 ∼12T, medium-sized, R0 = 1.85m, tokamak that is presently under construction in Devens, MA, led by Commonwealth Fusion Systems. It will be used to de-risk the high-field tokamak path to a fusion power plant and demonstrate the commercial viability of fusion energy. SPARC's first campaign plan is to achieve Qfus > 1 using an ICRF-heated, <10MW, high current, Ip ∼ 8.5 MA, L-mode fueled by D-T gas injection, and its second campaign will investigate H-mode operations in D-D. To facilitate plasma control and scientific learning, a targeted set of ∼50 plasma diagnostics are being designed and built for operation during these campaigns. While nearly all diagnostics are based on established techniques, the pace of deployment, relative to the first plasma, and the harshness of the thermal, electromagnetic, and radiation environment are unprecedented for medium-sized tokamaks. An overview of the SPARC diagnostic set is given, providing context to further details communicated by the SPARC team in companion publications that are system-specific. The system engineering philosophy for SPARC diagnostics is outlined, and the design and engineering verification process for components inside and outside the primary vacuum boundary are described. Diagnostics are mounted directly to the vacuum vessel as well as housed within a series of eight midplane and 24 off-midplane replaceable port plugs. With limited exceptions, signal conditioning, digitization electronics and cameras as well as lasers and microwave sources are localized to a series of five Diagnostic Lab spaces, totaling ∼350m2, located >15m from the center of the tokamak, on the other side of a 2.4m concrete shielding wall. A series of 31 large-scale penetrations have been included in the SPARC Tokamak Hall to facilitate integration of early campaign diagnostics and to provide upgradability.

Development of the neutral gas diagnostic system for neutral pressure measurements and gas analysis on the SPARC tokamak.

A suite of plasma diagnostics will be installed on the SPARC tokamak to allow for real-time plasma control, an investigation of high-field tokamak physics, and to de-risk the design of ARC, a compact fusion power plant with the aim to supply electricity to the grid. Among these diagnostics is the neutral gas diagnostics system (NTGS), a set of pressure sensors and gas analyzers used to monitor neutral pressure and gas composition for plasma control, optimization of wall conditioning, and helium ash removal, among other measurement functions linked to operational and scientific research needs. While reliable measurements of neutral pressure and gas composition have been fielded on existing magnetic-confinement fusion devices, SPARC represents a step increase in challenge due to its larger power density, higher field, high vacuum vessel bake temperatures, and higher neutron flux environment, as well as a step decrease in the accessibility for maintenance of in-vessel sensors. Multiple sensor types will be employed to have defense-in-depth and mitigate common failure modes. The NTGS system is currently progressing through final design, working to close out decisions using prototyping and analysis, and then moving on to procuring sensors for assembly and installation on SPARC. This paper outlines the current status of the system design and the diagnostic requirements that motivate neutral gas measurements on SPARC, as well as highlights the planned prototyping activities.

Safety Analysis of Metallic Tokamak-I Against Electromagnetic and Structural Loads

During tokamak operation, the structural integrity of the vacuum vessel (VV) of Metallic Tokamak-I (MT-I), a small spherical tokamak, was evaluated. This evaluation involved simulating real experimental data of electromagnetic (EM) and structural loads using the ANSYS platform. Internal heat generation, induced currents, and inertial and pressure loads in the VV were analyzed to determine their effects on the VV. This analysis was conducted on a 180-deg sector model over a 10-ms-event period. To create multiple checkpoint events, the plasma current was assumed to be formed at variable positions of the VV, hence inducing variable current for each event. The events are divided into four cases based on the radial and vertical displacements of plasma. The response of the VV structure was calculated using coupling of EM and structural modules of ANSYS. It is observed from the numerical results that the maximum stress on the VV is in a safe range and that the temperature rise on the vessel can be reduced by natural convection only if the event is ended in 10 ms. A prolonged event can result in permanent deformation in the VV structure. A disruption event on the limiter region is also studied.

Reduced order modeling for real-time monitoring of structural displacements due to electromagnetic forces in large scale tokamaks

Abstract The real-time monitoring of the structural displacement of the vacuum vessel of thermonuclear fusion devices caused by electromagnetic loads is of great interest. In this paper, model order reduction is applied to the integral equation methods and the finite elements method to develop electromagnetic and structural reduced order models (ROMs) compatible with real-time execution which allows for the real-time monitoring of strain and displacement in critical positions of Tokamaks machines. Low-rank compression techniques based on hierarchical matrices are applied to reduce the computational cost during the offline stage when the ROMs are constructed. Numerical results show the accuracy of the approach and demonstrate the compatibility with real-time execution in standard hardware.

Performances of the ITER Pressure Suppression System during unstable steam condensation regimes

Abstract The paper deals with the qualification of the main safety system of the nuclear fusion reactor ITER: the Vacuum Vessel Pressure Suppression System (VVPSS). This safety system manages the Loss of Coolant Accident (LOCA), that could occur in the Vacuum Vessel (VV), avoiding a significant increase of the internal pressure that could breach the primary confinement barrier. Steam condensation occurs at sub-atmospheric pressure. This condition is original respect the usual applications of steam direct condensation in nuclear fission reactors. An extensive experimental research program, funded by ITER Organization, on reduced scale experimental rigs (scale 1/22 and 1/10) have been carried out at the University of Pisa. Unstable condensation regimes occurred during some tests at about 500 g/s of steam mass flow rate (10 % of the maximum steam mass flow rate foreseen for the relevant accidental scenarios). In these conditions, high intensity vibrations of the whole experimental rig occurred. These events can be classified as ‘Chugging’ or Condensation Induced Water Hammer (CIWH). &amp;#xD;The performed tests have been analysed elaborating the images of the video-cameras located in the pool and comparing the steam jet shape with the acceleration values recorded by an accelerometer mounted on the sparger structure.&amp;#xD;These elaborations permitted to evaluate the dynamic of bubble collapse, the volume of the developed external bubbles, the collapse frequency as function of thermal-hydraulic parameters. &amp;#xD;Moreover, the paper emphasizes the beneficial effects of non-condensable gases on the unstable condensation regimes. In fact, the non-condensable gas inhibits the occurrences of Water Hammer inside the sparger and eliminates the thermal stratification inside the water pool increasing the turbulence.&amp;#xD;

Preliminary compact torus injection experiments on the Keda Torus eXperiment reversed field pinch

A new compact torus injector (KTX-CTI) has been built for injection experiments on the Keda Torus eXperiment (KTX) reversed field pinch (RFP). The aim is to study the fundamental physics governing the compact torus (CT) central fueling processes. In experiments conducted under the sole influence of a 0.1 T toroidal magnetic field, the injected CT successfully penetrated the entire toroidal magnetic field, reaching the inner wall of the KTX vacuum vessel. Upon reaching the inner wall, the CT diffused both radially outward and toroidally within the vessel at a discernible diffusion speed. Moreover, the inherent helicity within the CT induced a modest KTX plasma current of 200 A, consistent with predictions based on helicity conservation. CT injection demonstrated the capability to initiate KTX discharges at low loop voltages, suggesting its potential as a pre-ionization and current startup technique. During RFP discharges featuring CT injection, the central plasma density was found to exceed the Greenwald density limit, with more peaked density profiles, indicating the predominant confinement of CT plasma within the core region of the KTX bulk plasma.

ITER materials irradiation within the D–T neutron environment at JET: post-irradiation radioactivity analysis following the DTE2 experimental campaign

Abstract This work presents the results following the first irradiation of ITER materials samples in a tokamak D--T plasma environment operating at significant fusion power. The materials exposed to this nuclear environment at the Joint European Torus during the DTE2 experimental campaign that took place in 2021 include representative ITER samples from various components such as poloidal field (PF) coil jacket samples, toroidal field coil radial closure plate steels, EUROFER 97 steel, W and CuCrZr materials from the divertor, Inconel-718, CuCrZr, and 316L stainless steel for blanket modules, as well as vacuum vessel forging samples.&amp;#xD;&amp;#xD;The experimental results discussed include high-resolution gamma spectrometry measurements and analysis conducted with the post-irradiated samples, of which there were 68 in total. These samples were exposed through different experimental campaigns, including deuterium, deuterium--tritium and tritium phases. Supporting the analysis were 25 dosimetry foil-based neutron diagnostics and two 'VERDI' neutron spectrometry diagnostics. A further 12 samples for positron annihilation spectroscopy (PALS) were also irradiated. The irradiation of all these samples took place in a long-term irradiation assembly located near the JET vacuum vessel.&amp;#xD;&amp;#xD;The post-irradiation analysis of the ITER material samples has yielded valuable insights into their material activation levels and radiation fields. Comparative assessments between experimental measurements and comprehensive neutronics simulations have demonstrated a significant level of agreement in this work, while also revealing some discrepancies in specific material instances. The data and interpretation from this work not only serve as a robust experimental foundation for enhancing the precision and predictability of neutronics simulation approaches for ITER and next-step devices but also present some opportunities for the refinement of simulation methodologies. In light of these findings, a series of recommendations have been proposed, aimed at improving confidence in nuclear predictions associated with materials that have been exposed to fusion nuclear environments and advancing understanding in this important domain.

Divertor Tokamak Test facility project: status of design and implementation

Abstract An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, XD divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

Wavelet denoising analysis on vacuum-process monitoring signals of aerospace vacuum vessel structures

Abstract The monitoring of micro-defects or external actions induced vacuum degradation in aerospace vacuum vessels is an important challenge. A vacuum-process monitoring method based on quasi-distributed fiber Bragg grating (FBG) sensing technology is proposed. Due to the influence of environmental noise and vacuum pump operation noise, the raw signals measured by FBG sensors contain a large amount of noises, which affects the measurement accuracy and data analysis. Therefore, the wavelet threshold (WT) denoising method is proposed to analyze the influence of noises on the monitoring signals measured by two different kinds of FBG sensors. The evaluation of the monitoring signals after denoising indicates that the proposed method can effectively remove the noise and significantly improve signal quality. The highest signal-to-noise ratio of the processed signals can reach 37.61 dB and the mean square error is 3.68 × 10−7, while retaining the key features of the original signal. The proposed WT denoising method demonstrates better performance and feasibility compared with moving average filtering and Kalman filtering methods. The study provides critical supports for improving the performance and reliability of the vacuum vessel monitoring system.

Effect of gamma-radiation on the transmission and return losses of FC/APC connectors

The Fibre Optics Current Sensor (FOCS) installed at ITER will use optical fibres to perform plasma current measurements based on the Faraday effect. A connection between the sensing fibre placed on the Vacuum Vessel and the fibre guiding the signal to the data-acquisition hardware will be required. Because of practical reasons, this link contains FC/APC connectors, which guarantee the low return loss required for normal operation of the FOCS. The connectors will be located in the port cell area and will be subject to high levels of nuclear radiation. Radiation may influence both transmission and return losses of the FC/APC connectors and, as a result, degrade the FOCS performance. To address the problem, several FC/APC connectors were exposed to Co60 gamma radiation up to a 550 kGy dose. Transmission through the connectors was monitored during irradiation. The return loss of mated FC/APC connectors was measured before and after irradiation. It was found that radiation-induced changes of both transmission and reflection parameters are acceptable for the FOCS at ITER.

Realization of a gas puff imaging system on the Wendelstein 7-X stellarator.

A system for studying the spatiotemporal dynamics of fluctuations in the boundary of the W7-X plasma using the "Gas-Puff Imaging" (GPI) technique has been designed, constructed, installed, and operated. This GPI system addresses a number of challenges specific to long-pulse superconducting devices, such as W7-X, including the long distance between the plasma and the vacuum vessel wall, the long distance between the plasma and diagnostic ports, the range of last closed flux surface (LCFS) locations for different magnetic configurations in W7-X, and management of heat loads on the system's plasma-facing components. The system features a pair of "converging-diverging" nozzles for partially collimating the gas puffed locally ≈135mm radially outboard of the plasma boundary, a pop-up turning mirror for viewing the gas puff emission from the side (which also acts as a shutter for the re-entrant vacuum window), and a high-throughput optical system that collects visible emission resulting from the interaction between the puffed gas and the plasma and directs it along a water-cooled re-entrant tube directly onto the 8 × 16 pixel detector array of the fast camera. The DEGAS 2 neutral code was used to simulate the Hα (656nm) and HeI (587nm) line emission expected from well-characterized gas-puffs of H2 and He and excited within typical edge plasma profiles in W7-X, thereby predicting line brightnesses used to reduce the risks associated with system sensitivity and placement of the field of view. Operation of GPI on W7-X shows excellent signal-to-noise ratios (>100 at 2 Mframes/s) over the field of view for minimally perturbing gas puffs. The GPI system provides detailed measurements of the two-dimensional (radial and poloidal) dynamics of plasma fluctuations in the W7-X edge and scrape-off layer and in and around the magnetic islands outside the LCFS that make up the island divertor configuration employed on W7-X.

Unlocking maintenance-architecting STEP for maintenance and realizing remountable magnet joints.

The architecture of the Spherical Tokamak for Energy Production (STEP) has been developed to enable a hybrid maintenance approach using ports in the vacuum vessel for a limited list of tasks that must be performed shortly after shutdown, and larger openings to simplify and speed up major refits. Robotic handling systems in zero-human entry facilities will prevent workers from being exposed to the most hazardous environments. While the approach is largely grounded in existing technologies, the scale and environment of STEP will require significant technology development. Notably, programmes have been established to develop service connections and in-vessel robotic technologies. The engineering integration of the maintenance strategy into the tokamak remains a priority, as does ongoing work to simplify and reduce the cost of the buildings required to facilitate maintenance. Remountable magnet joints are critical to ensuring life-limited magnet components can be replaced during the STEP lifetime and realizing the STEP maintenance strategy. It is a high-risk endeavour owing to the low technology maturity of the potential solutions and owing to the tough and intertwined technical challenges and constraints imposed by both the fundamental physics and the STEP requirements and architecture. An integrated design approach has been taken to balance many competing factors and integrate with interfacing systems, and a multi-faceted technology development programme has been established to address technical risk and to inform, verify and validate the STEP remountable magnet design. This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

Development of double-walled bellows for ITER VUV spectrometer systems

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.

Design and development of FPGA based trigger system for automation of metallic tokamak (MT-I)

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.

Design and experiment results of HL-3 baking system

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.