Fusion Engineering and Design最新文献

{"title":"Tungsten coated tiles for KSTAR PFC upgrade","authors":"Hee-Jae Ahn ,&nbsp;Hyun-Ki Park ,&nbsp;Koung Moon Kim ,&nbsp;Kwang Pyo Kim ,&nbsp;Kaprai Park ,&nbsp;Sungjin Kwon ,&nbsp;Young Min Park ,&nbsp;E. Grigore ,&nbsp;Si-Woo Yoon","doi":"10.1016/j.fusengdes.2025.115055","DOIUrl":"10.1016/j.fusengdes.2025.115055","url":null,"abstract":"<div><div>The KSTAR plasma facing components (PFCs) include graphite tiles and metal back-plates. Since their installation in 2010, the PFCs have effectively fulfilled their mission. KSTAR's PFCs are being upgraded to create a fully tungsten (W) environment, designed to handle the increased thermal loads associated with maximum heating power and extended pulse lengths. Tungsten's high melting temperature, its heaviness, brittleness, and costliness led to the consideration of W-coated carbon tiles as a viable compromise for the limiters and passive stabilizer. The newly designed inboard limiter tiles feature a W-coated rounded surface facing the plasma, twice the size to mitigate leading-edge effects. Standard tile types were minimized to save on manufacturing and maintenance costs. Approximately 1500 graphite tiles were newly manufactured due to surface and radiation contamination. These tiles were coated with 13 μm W and 2 μm Mo using Romania's NILPRP PVD CMSII, a technique also applied to JET, ASDEX-U, and WEST PFCs. The Mo layer enhances adhesion by addressing thermal expansion mismatches between graphite and W. The W coating was completed in early 2024, and the installation of W-coated tiles is scheduled before the 2026 campaign.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115055"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808139","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Microwave stray radiation measurement techniques","authors":"J.W. Oosterbeek ,&nbsp;M. Stern ,&nbsp;K.J. Brunner ,&nbsp;M. Hirsch ,&nbsp;B. Kursinski ,&nbsp;H.P. Laqua ,&nbsp;R.J. Zubieta-Lupo,&nbsp;S. Marsen ,&nbsp;A. Moro ,&nbsp;D. Moseev ,&nbsp;S. Pak ,&nbsp;A. Sirinelli ,&nbsp;C. Sozzi ,&nbsp;T. Stange ,&nbsp;R.C. Wolf ,&nbsp;W7-X Team","doi":"10.1016/j.fusengdes.2025.114967","DOIUrl":"10.1016/j.fusengdes.2025.114967","url":null,"abstract":"<div><div>In magnetic-confinement fusion experiments, such as tokamaks and stellarators, microwave stray radiation loads occur due to non-absorbed electron cyclotron heating power, or – at very high electron temperatures – due to cyclotron emission from the fusion plasma itself. The metallic first wall of the plasma vessel reflects most of the non-absorbed power and generates a more or less homogeneous stray radiation field, which causes ohmic and dielectric heating in lossy components, while in microwave receivers outside the vessel it may cause overload or even destruction of mixers and detectors. These issues are understood and mitigation measures are developed or already in place. However, accurate and fast diagnoses of stray radiation provides a challenge: stray radiation may not be a truly scrambled mixture of k-vectors and polarisation. In such case, the signal of a single-mode microwave detector will show strong fluctuations in time. Bolometers, with a size such that many modes are simultaneously measured, provide a much more stable signal due to the integration, but the time response is of the order of seconds. And an inherent aspect of bolometers is that at elevated temperatures they become non-linear due to radiative heat losses. A caloric measurement, using for instance a water-cooled microwave load, gives a calibrated and reproducible measurement but is very slow — typically several minutes. Due to size and water coolant, calorimetry is challenging to integrate into a fusion experiment. However, it is useful for cross-calibration of bolometers. In this paper microwave stray radiation is briefly introduced, then followed by an overview of measurement techniques in use at the Wendelstein 7-X stellarator (W7-X) as well as development work together with ITER.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114967"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808209","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Innovative RAFM steels with improved impact properties","authors":"Carlo Cristalli ,&nbsp;Luciano Bozzetto ,&nbsp;Alessandro Colaneri ,&nbsp;Roberto Sorci ,&nbsp;Oriana Tassa","doi":"10.1016/j.fusengdes.2025.115034","DOIUrl":"10.1016/j.fusengdes.2025.115034","url":null,"abstract":"<div><div>The contribution of ENEA together with Rina-CSM to the Eurofusion programme “WPMAT\" (Work Package MATerials) deals with the development of innovative RAFM steels able to withstand the critical low temperature (280–300 °C in the “water-cooling” option) foreseen for the blanket of the first DEMO reactor. Therefore the target is the development of tougher alloys, suitable to prevent the low temperature irradiation embrittlement. In addition to the DBTT shift due to the dpa damage under irradiation, also the issue of the increased embrittlement due to Helium production must be taken into account.</div><div>The alloy design focused on chemical composition optimization in terms of contents of some main key elements as Carbon, Manganese, Vanadium, Nitrogen and Tantalum. An overall indication deriving from the last neutron irradiation experiment was to increase Vanadium and Nitrogen at the expense of Carbon and Manganese. The ingot has then been cast with accurate control of impurities. The final product has proven compliant to the specifications concerning RAFM steels for fusion environment.</div><div>Alternative thermo-mechanical treatments have then been applied on this RAFM steel. On the one hand the DBTT after a standard double normalization and tempering treatment is definitely successful; -140 °C on KLST specimens. The extremely low DBTT must be considered a consequence of the choices done on the chemical composition. On the other hand a particular TMT, whose aim is to induce recrystallization of the laths substructure and obtain an equiaxed nano-grain structure, has been investigated. Supposed to increase the resistance to crack propagation, the decrease in grain size (less than one μm) leads to a transition curve that doesn't show a sharp drop in the absorbed energy, but instead a more gradual transition from the USE to the LSE. The tensile properties in the recrystallized condition result markedly higher than after the conventional treatment.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115034"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808212","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental validation of transmutation products calculations in neutron irradiated tungsten","authors":"V. Chatzikos ,&nbsp;M.I. Savva ,&nbsp;T. Vasilopoulou ,&nbsp;I.E. Stamatelatos ,&nbsp;D. Terentyev ,&nbsp;A. Stankovskiy ,&nbsp;N. Patronis ,&nbsp;K. Mergia","doi":"10.1016/j.fusengdes.2025.115012","DOIUrl":"10.1016/j.fusengdes.2025.115012","url":null,"abstract":"<div><div>Neutron irradiation can significantly alter the elemental composition of a material by generating transmutation products through nuclear reactions between neutrons and atomic nuclei. These changes in composition can substantially impact the material's physical and mechanical properties. Therefore, accurately assessing the buildup of transmutation products in neutron-irradiated materials is essential for understanding and predicting these effects. Tungsten (W) is particularly critical for the first wall and divertor components in fusion reactors. As such, an accurate assessment of transmutation products in neutron-irradiated tungsten is crucial for the safety and lifespan of future fusion power plants. The scope of the present work is to experimentally validate calculations of transmutation products buildup in tungsten after neutron irradiation at the Materials Test Reactor (MTR) BR-2 at SCK CEN, Belgium. Tungsten specimens were irradiated to doses of 0.12 and 0.19 displacements per atom (dpa) within the temperature range of 600 to 1200 °C. Nuclide inventory calculations were performed using the FISPACT-II code and the TENDL-2019 cross-section library. <em>γ</em>-ray spectroscopy was employed to determine the specific activity levels of <span><math><mrow><msup><mrow></mrow><mn>181</mn></msup><mi>W</mi></mrow></math></span>, <span><math><mrow><msup><mrow></mrow><mn>185</mn></msup><mi>W</mi></mrow></math></span>, <span><math><mrow><msup><mrow></mrow><mn>188</mn></msup><mi>W</mi></mrow></math></span>, <span><math><mrow><msup><mrow></mrow><mn>188</mn></msup><mi>R</mi><mi>e</mi></mrow></math></span> and <span><math><mrow><msup><mrow></mrow><mn>182</mn></msup><mi>T</mi><mi>a</mi></mrow></math></span> in order to validate the transmutation products calculations for rhenium (Re), osmium (Os) and tantalum (Ta). It is shown that the theoretical calculations for Re and Os concentrations are in good agreement with the experimental data, while the Ta concentration is underestimated by a factor of approximately 1.5.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115012"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808136","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Remote handling validation of the OPM-FDS connecting system for the IFMIF-DONES target assembly","authors":"G. Benzoni ,&nbsp;D. Bernardi ,&nbsp;D. Fuentes-Calero ,&nbsp;A. Cammi ,&nbsp;V. Claps ,&nbsp;C. Introini ,&nbsp;G. Miccichè ,&nbsp;F.S. Nitti ,&nbsp;D. Sánchez-Herranz ,&nbsp;R. Mozzillo ,&nbsp;C. Tripodo","doi":"10.1016/j.fusengdes.2025.115032","DOIUrl":"10.1016/j.fusengdes.2025.115032","url":null,"abstract":"<div><div>The qualification process of the materials for future DEMO fusion reactors plays a pivotal role towards its realization. In this light, the International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source (IFMIF-DONES) has achieved the final design stage and is currently under construction in Granada (Spain). IFMIF-DONES aims to reproduce the irradiation conditions foreseen on the first wall of DEMO by using deuteron-lithium stripping reactions induced by a D+ beam accelerated through a superconducting LINAC. These reactions occur in a Target Assembly (TA) within a closed cavity (Test Cell) housing the test modules. Due to the strong neutron activation, a Remote Handling System (RHS) is required to perform the maintenance operations in the Test Cell. Therefore, suitably custom-designed components necessary due to the conditions of IFMIF-DONES, must be validated from the RH point of view. In particular, the system responsible for connecting the TA with the accelerator beamline plays a key role. This system includes a compression element made of an extensible bellow actuated by a One-Point-Mechanism (OPM) and a circular chain acting as a Fast Detachment System (FDS), which secures the connection by applying the required tightening force on the sealing gasket. The main objective of this work is then to test and validate the OPM-FDS design with focus on the RH suitability. Therefore, to demonstrate the concept validity and qualify the system, the University of Granada made a prototype of the OPM-FDS connection in collaboration with ARQUIMEA® and delivered it to the ENEA Brasimone Center for experimental validation. The outcomes of the validation activity conducted in this work confirmed that the OPM-FDS design is suitable for Remote Handling and can perform several connections and disconnections. The rescue system, which detaches the FDS in case of failure, has also been tested successfully. The experimental campaign also showed some potential improvements in the design of the connecting system for further implementation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115032"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808137","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"MaPLE: A versatile facility for investigating MHD convective flows","authors":"C. Koehly,&nbsp;C. Courtessole,&nbsp;L. Bühler","doi":"10.1016/j.fusengdes.2025.115031","DOIUrl":"10.1016/j.fusengdes.2025.115031","url":null,"abstract":"<div><div>With MaPLE (<u>Ma</u>gnetohydrodynamic <u>P</u>b<u>L</u>i <u>E</u>xperiment), KIT provides an experimental research facility to investigate magnetohydrodynamic (MHD) effects affecting liquid metal blankets of future fusion reactors using the prototypical liquid metal PbLi. The facility is designed to enable heat transfer studies in free, forced, or mixed magneto-convective flows in simple ducts and more complex typical geometries relevant to liquid metal breeder blankets. It features a versatile infrastructure that allows experimental examination of flows at any orientation with respect to gravity under a transverse magnetic field, as those occurring in a fusion blanket. While MaPLE has been primarily designed to explore the combined MHD and buoyancy effects, the facility can also support the development of measuring techniques at high temperatures and strong magnetic fields. In synergy with the well-established MEKKA facility, which offers an even larger experimental magnetic volume, MaPLE is also suited for analyzing PbLi flows in complex geometries such as scaled mock-ups of fusion blankets. Identified as one of EUROfusion Technological Hubs, MaPLE will support the development of breeding blankets the European demonstration fusion power plant DEMO, and the international ITER project. The main features and characteristics of the facility are described in this paper, and the roadmap for the first experimental campaigns is presented.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115031"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808217","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical investigation of the DTT cryopump performance via 3D Direct Simulation Monte Carlo modeling","authors":"C. Tantos ,&nbsp;H. Strobel ,&nbsp;V. Hauer ,&nbsp;C. Day ,&nbsp;T. Giegerich ,&nbsp;P. Innocente","doi":"10.1016/j.fusengdes.2025.115021","DOIUrl":"10.1016/j.fusengdes.2025.115021","url":null,"abstract":"<div><div>DTT (Divertor Tokamak Test Facility) is a new facility, currently under construction in Frascati, Italy. The goal is to provide a basis to perform various scaled experiments for testing different magnetic configurations and alternative solutions for the power exhaust system of DEMO. The DTT pumping system design uses cryopumps as the primary pumping solution and up to 10 pumping openings. The cryopump system was developed at the Karlsruhe Institute of Technology (KIT). In the present work, the pumping capabilities of the DTT cryopump are estimated by performing a 3D numerical investigation of the neutral gas dynamics in the pumping duct of DTT including the entire complex cryopump geometry. The investigation is based on the Direct Simulation Monte Carlo (DSMC) method, which allows for a precise description of the neutral gas dynamics over the entire range of the gas collisionality. The values of the pumping probability for deuterium and neon were determined for two scenarios: open and closed divertor toroidal gaps. For open gaps, the probabilities were obtained as 0.4 for deuterium and 0.62 for neon, with a slight increase in both values observed when the gaps were closed. The results suggest that achieving the target of ten ports for deuterium pumping seems feasible, while for neon, partial elimination of toroidal leakages may be required. The importance of these simulations lies in the fact that, given the imposed simplifications, they delineate the existing pumping capabilities of the DTT particle exhaust and can act as a guide regarding the pumping capabilities among various plasma configurations. Moreover, this work demonstrates the level of geometric complexity that can be adopted in numerical modeling and highlights the effort needed to determine the values of the pumping probability, which are dependent on the chosen pumping technology.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115021"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768180","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"On the use of LSTM-based estimation components for tokamak gas actuator control","authors":"Hudson Baker,&nbsp;Lucas Brown,&nbsp;Adam Parrott","doi":"10.1016/j.fusengdes.2025.114932","DOIUrl":"10.1016/j.fusengdes.2025.114932","url":null,"abstract":"<div><div>During the operation of JET, an approach using a Long Short-Term Memory (LSTM) neural network was explored to provide a data-driven means of developing and deploying a control system component. The LSTM was trained to model the piezo actuator valve which has non-linear hysteresis effects and required occasional recalibration due to drifts. A further motivation was to develop a process to make use of the large amount of control system data to enable control system performance improvements whilst minimising the use of machine time. The LSTM is developed in the form of a first-order Markov model where it takes inputs from pressure sensors around a gas reservoir and a control signal to the gas valve and predicts the next state of the downstream pressure sensor. A comparison of the LSTM time series model is done with two different linear Kalman filters.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114932"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768181","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Integrity assessment of Tokamak-type fusion reactor First Wall and Blanket structures","authors":"Mohammad Alabdullah ,&nbsp;Nasr Ghoniem","doi":"10.1016/j.fusengdes.2025.114995","DOIUrl":"10.1016/j.fusengdes.2025.114995","url":null,"abstract":"<div><div>This study presents a nonlinear structural analysis of the First Wall &amp; Blanket (FW/B) of the Fusion Nuclear Science Facility (FNSF), aimed at identifying potential failure modes and estimating the component’s lifetime. The analysis focuses on the accumulation of plastic strain damage and applies fracture mechanics principles to assess the effects of radiation on mechanical properties. The study uses a database for reduced activation ferritic/martensitic steel F82H, with particular emphasis on radiation-induced changes in its mechanical properties. Safety factors for various failure modes are evaluated both at the beginning of life and after neutron irradiation under normal and off-normal operating conditions. The results indicate that the most critical failure modes are flow localization, thermal creep damage, and fast fracture due to radiation-induced reduction in fracture toughness. Based on current empirical prediction methods, we estimate that the FW/B structure will likely survive for approximately one year of operation under conditions of 1.2 MW/m<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span> neutron wall loading and 0.25 MW/m<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span> surface plasma thermal flux. These findings suggest that the effects of very high fluence (100–200 dpa) may be less significant than expected, with thermal creep damage and rapid fracture toughness degradation being the primary factors controlling failure. Therefore, there is a critical need to improve creep strength at temperatures above 500 °C and to mitigate fracture toughness degradation with increasing neutron fluence. Furthermore, the development of compact fusion energy sources with neutron wall loads greater than 5 MW/m<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span> may require new classes of structural materials to ensure their economic viability.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114995"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768100","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development and optimization of a new control interface for the ASDEX upgrade high current converters","authors":"Friedhelm Albers,&nbsp;Nils Arden,&nbsp;Manuel Krause,&nbsp;Bernhard Sieglin,&nbsp;the ASDEX Upgrade Team","doi":"10.1016/j.fusengdes.2025.115006","DOIUrl":"10.1016/j.fusengdes.2025.115006","url":null,"abstract":"<div><div>The original interface between the ASDEX Upgrade (AUG) control system and the power converters of the high current group uses multicore signal cables to transmit command, status and control loop signals. Depending on the configuration of the AUG coils and the converters, these signal cables were manually arranged via connectors on a patch panel.</div><div>The redesigned control interface will consist of two separate Ethernet links, one for command and status signals and one for control loop signals. With the digital interface, there are no more contacting problems or crosstalk of signals.</div><div>A new programmable logic controller (PLC) was installed as the central point of the network to coordinate the exchange of command and status signals between AUG and the power converters. In order to achieve short response times in the network, additional PLC hardware was replaced and the real-time capable network protocol PROFINET IO was selected.</div><div>The second Ethernet connection, which uses the User Datagram Protocol (UDP), is currently under development. Therefore, an industrial PC was installed to receive the data from AUG and forward it to the corresponding converter feedback control system. While the newer converters with digital feedback control can be connected directly via Ethernet, the older converters with analog feedback control require additional digital-to-analog converters.</div><div>This paper presents the concept, implementation and optimization of the command and status interface. Furthermore, the design of the prototype of the control loop interface for the converters with analog feedback control as well as initial test results are shown.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115006"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768179","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}