Fusion Engineering and Design最新文献

{"title":"The tokamak system code D0FUS and its first applications: Impact of B and confinement scaling law on power plant design","authors":"Timothé Auclair,&nbsp;Eric Nardon,&nbsp;Yanick Sarazin,&nbsp;Jean-François Artaud,&nbsp;Baptiste Boudes,&nbsp;Clarisse Bourdelle,&nbsp;Jean-Luc Duchateau,&nbsp;Alexandre Torre","doi":"10.1016/j.fusengdes.2025.115270","DOIUrl":"10.1016/j.fusengdes.2025.115270","url":null,"abstract":"<div><div>D0FUS is a new system code written in Python and designed to enable fast, extensive, and easily interpretable parameter scans. Inspired by the approach of Freidberg et al. (2015), D0FUS computes the main plasma stability indicators as well as the associated radial build. The code’s architecture and capabilities are reviewed, and its governing equations are presented. First results show possible viable designs at large aspect ratio, explore possible operations at high magnetic field and highlight the critical role of the energy confinement time scaling laws.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115270"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144522837","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":"A fuzzy-based approach to protect the steering mechanism of high-power antennas in Tokamak","authors":"Mehdi Davoudi ,&nbsp;Mohsen Davoudi","doi":"10.1016/j.fusengdes.2025.115290","DOIUrl":"10.1016/j.fusengdes.2025.115290","url":null,"abstract":"<div><div>When a mechanical actuator is located in an antagonistic environment where there is no possibility to place sensors to measure its position, there is the necessity of employment of a model-predictive protection system to stop the actuator to avoid danger in case of the malfunctioning actuator controller. In fact, due to the complexity of the actuator dynamics and its workspace map, it is not possible to use limiting proximity estimators or some controller output thresholds. This paper has aimed to design and implement a reliable protection strategy that monitors the movements of the actuator in real-time to protect the actuator from touching the workspace border especially when the actuator moves at its highest speed. The idea of the fuzzy-based scheme is to convert the workspace boundary map into safety contours using image processing and develop a model of the actuator with emergency braking using the Sugeno-type fuzzy inference system and Mamdani fuzzy system. The workspace information and dynamic model of the actuator and brake have been incorporated into the fuzzy model which is then being optimized using Adaptive Neuro-Fuzzy Inference System (ANFIS) to increase the accuracy. The fuzzy protection system raises the problem of the conventional model predictive-based protection system and its demand for high computational power and insufficient refresh rate. The proposed method can be applied to a variety of applications but the chosen one is for the protection of steering microwave antennas placed inside a nuclear fusion reactor for comparison of performance.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115290"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144522838","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":"A laser-based diagnostic for in situ monitoring of fuel retention in ITER","authors":"A. Huber ,&nbsp;Ph. Andrew ,&nbsp;G. Sergienko ,&nbsp;J. Assmann ,&nbsp;D. Castano ,&nbsp;A. De Schepper ,&nbsp;S. Friese ,&nbsp;I. Ivashov ,&nbsp;D. Kampf ,&nbsp;Y. Krasikov ,&nbsp;H.T. Lambertz ,&nbsp;Ph. Mertens ,&nbsp;K. Mlynczak ,&nbsp;K. Rasinska ,&nbsp;M. Schrader ,&nbsp;D. Van Staden ,&nbsp;A. Terra ,&nbsp;Xi Jiang ,&nbsp;M. Zlobinski ,&nbsp;A. Reutlinger ,&nbsp;Ch. Linsmeier","doi":"10.1016/j.fusengdes.2025.115298","DOIUrl":"10.1016/j.fusengdes.2025.115298","url":null,"abstract":"<div><div>This paper addresses the critical challenge of tritium inventory control in ITER and future fusion devices, emphasizing the necessity for precise measurement and spatial distribution of tritium within the vacuum vessel. The proposed laser-based T-monitor diagnostic system from Forschungszentrum Jülich employs Laser-Induced Desorption (LID) combined with Residual Gas Analysis (RGA) to measure hydrogen isotope concentrations on the inner divertor tiles. Key design elements include high-power laser integration, advanced optical systems, and a Fast Scanning Mirror Unit for accurate laser spot positioning. The diagnostic aims to measure <em>in situ</em> tritium retention, improving operational safety and efficiency in nuclear fusion environments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115298"},"PeriodicalIF":1.9,"publicationDate":"2025-06-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144510882","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":"Structural analysis of the US ITER central solenoid magnet cryogenic piping system","authors":"Aravind Shanmugasundaram ,&nbsp;Kevin Freudenberg ,&nbsp;Eric Morris ,&nbsp;Joshua Rathbun ,&nbsp;Robert L. Myatt ,&nbsp;Kristine B. Cochran ,&nbsp;Alex Istomin","doi":"10.1016/j.fusengdes.2025.115291","DOIUrl":"10.1016/j.fusengdes.2025.115291","url":null,"abstract":"<div><div>The Central Solenoid (CS) is the heart of the ITER tokamak and serves as a critical element in the ITER magnet system. The CS consists of a stack of six independently operated high field superconducting magnets (modules), a pre-compression structure, a support system, and a centering system.</div><div>The CS piping system supplies the modules and support structure with supercritical helium (He) coolant to achieve and maintain the 4 K operating temperature. The CS module piping consists of a system of inlet and outlet pipes that delivers the coolant to the modules and returns it to the cryostat. The CS structure piping supplies coolant from the structure cooling feeder to CS structure components such as key blocks, load distribution plates, and tie plates. The loads on the piping system and its support structure include dead load, internal coolant pressure, seismic accelerations, displacements induced by CS thermal contraction, displacements from electromagnetic (EM) effects on the CS components, and temperature differentials during cooldown and quench.</div><div>Four different ANSYS® static stress and dynamic modal models apply these loads to determine design-basis stresses and displacements which are used to evaluate the adequacy of the piping system. The piping and support components pass the static stress and fatigue requirements for the respective function and material of each component based on temperature-dependent stress limits. Modal analysis indicates the natural frequency of all the CS cooling system piping lies between 4 Hz and 10 Hz.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115291"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144510881","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":"Engineering considerations for optical diagnostics of European DEMO","authors":"I. Katona ,&nbsp;D. Dunai","doi":"10.1016/j.fusengdes.2025.115292","DOIUrl":"10.1016/j.fusengdes.2025.115292","url":null,"abstract":"<div><div>Optical diagnostics measurements were identified as necessary for plasma control in the initial concept for the plasma diagnostic and control (D&amp;C) system of European DEMO. The protection of the first mirror is identified as a critical subject for future large sized fusion devices. Radio Frequency Plasma Discharge mirror cleaning solutions proposed for ITER have been discarded for fusion power plant environment. Erosion and deposition processes can be mitigated by neutral gas between the aperture and the front mirror. Currently 3 optical diagnostics are being developed for EU-DEMO. The Divertor Monitoring Diagnostic had the highest priority thus that was selected for conceptual development. Three optomechanical layout configurations were analyzed, with the optimal candidate further refined for improved integration and performance. The design emphasized the optical duct geometry, revealing that a shorter duct simplifies mechanical integration. Preliminary Monte Carlo N-Particle (MCNP) neutronics simulations estimated the diagnostic module's nuclear load, providing key insights for further refinement. This iterative workflow enables systematic optimization of the optical and mechanical structure to meet stakeholder requirements, yielding a robust, integrated diagnostic module design. The methodology offers a transferable framework for other optical diagnostics under development for EU-DEMO and future fusion power plants.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115292"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144501750","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":"Achieving high amplifications in a cw-driven optical cavity relevant for photoneutralization of negative ion beams","authors":"R. Friedl ,&nbsp;R. Borkenhagen ,&nbsp;U. Fantz","doi":"10.1016/j.fusengdes.2025.115244","DOIUrl":"10.1016/j.fusengdes.2025.115244","url":null,"abstract":"<div><div>In order to overcome the drawback of the limited neutralization efficiency of a gas neutralizer, laser neutralization is discussed for negative ion based neutral beam injection systems for future fusion devices. Those could easily deliver neutralization efficiencies far beyond the 60% constraint of conventional systems, in principle only limited by the provided optical power. For fusion-relevant ion beams, optical powers beyond tens of MW would be required, which poses the challenge of reliably providing such high powers in cw operation. Measures to amplify the optical power are thus mandatory and high-finesse optical cavities wrapped around the ion beam can be applied for this purpose. The target is to reduce the initially required laser power by three to four orders of magnitude and maintain stable coupling of the laser and the enhancement cavity in resonance. In order to confirm that such high amplifications can be reliably obtained in cw, an independent optical test bench was set up. The setup proved that using a two-fold feedback system (Pound–Drever–Hall locking scheme), stable resonance locking of laser and cavity for more than an hour is feasible. Furthermore, amplifications of 7’000 could be achieved experimentally, reaching in-cavity powers in the range of several kW, driven by only 240<!--> <!-->mW input power. Approaches to integrate such a laser neutralizer setup at the negative ion beam facility Batman Upgrade and options for cavity folding are discussed in terms of the achievable neutralized fraction of the ion beam.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115244"},"PeriodicalIF":1.9,"publicationDate":"2025-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144490407","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":"Electromagnetic system conceptual design for a negative triangularity tokamak","authors":"S. Guizzo ,&nbsp;M.A. Drabinskiy ,&nbsp;C. Hansen ,&nbsp;A.G. Kachkin ,&nbsp;E.N. Khairutdinov ,&nbsp;A.O. Nelson ,&nbsp;M.R. Nurgaliev ,&nbsp;M. Pharr ,&nbsp;G.F. Subbotin ,&nbsp;C. Paz-Soldan","doi":"10.1016/j.fusengdes.2025.115257","DOIUrl":"10.1016/j.fusengdes.2025.115257","url":null,"abstract":"<div><div>Negative triangularity (NT) tokamak configurations have several key benefits including sufficient core confinement, improved power handling, and reduced edge pressure gradients that allow for edge-localized mode (ELM) free operation. We present the design of a compact NT device for testing sophisticated simulation and control software, with the aim of demonstrating NT controllability and informing power plant operation. The TokaMaker code is used to develop the basic electromagnetic system of the <span><math><msub><mrow><mi>R</mi></mrow><mrow><mn>0</mn></mrow></msub></math></span> = 1 m, <span><math><mi>a</mi></math></span> = 0.27 m, <span><math><msub><mrow><mi>B</mi></mrow><mrow><mi>t</mi></mrow></msub></math></span> = 3 T, <span><math><msub><mrow><mi>I</mi></mrow><mrow><mi>p</mi></mrow></msub></math></span> = 0.75 MA tokamak. The proposed design utilizes eight poloidal field coils with maximum currents of 1 MA to achieve a wide range of plasma geometries with <span><math><mrow><mo>−</mo><mn>0</mn><mo>.</mo><mn>7</mn><mo>&lt;</mo><mi>δ</mi><mo>&lt;</mo><mo>−</mo><mn>0</mn><mo>.</mo><mn>3</mn></mrow></math></span> and <span><math><mrow><mn>1</mn><mo>.</mo><mn>5</mn><mo>&lt;</mo><mi>κ</mi><mo>&lt;</mo><mn>1</mn><mo>.</mo><mn>9</mn></mrow></math></span>. Scenarios with strong negative triangularity and high elongation are particularly susceptible to vertical instability, necessitating the inclusion of high-field side and/or low-field side passive stabilizing plates which together reduce vertical instability growth rates by <span><math><mo>≈</mo></math></span>75%. Upper limits for the forces on poloidal and toroidal field coils are predicted and mechanical loads on passive structures during current quench events are assessed. The 3 T on-axis toroidal field is achieved with 16 demountable copper toroidal field coils, allowing for easy maintenance of the vacuum vessel and poloidal field coils. This pre-conceptual design study demonstrates that the key capabilities required of a dedicated NT tokamak experiment can be realized with existing copper magnet technologies.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144489506","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":"ITER-relevant experimental neutronic activities at JET during DTE3 and at the Frascati neutron generator","authors":"N. Fonnesu ,&nbsp;P. Beaumont ,&nbsp;T. Berry ,&nbsp;A. Colangeli ,&nbsp;F. Dacquait ,&nbsp;M. Damiano ,&nbsp;D. Flammini ,&nbsp;C.L. Grove ,&nbsp;X. Litaudon ,&nbsp;S. Loreti ,&nbsp;M. Lungaroni ,&nbsp;S. Mianowski ,&nbsp;F. Moro ,&nbsp;S. Noce ,&nbsp;J. Peric ,&nbsp;A. Previti ,&nbsp;V. Radulović ,&nbsp;R. Villari ,&nbsp;P. Zito ,&nbsp;JET Contributors","doi":"10.1016/j.fusengdes.2025.115297","DOIUrl":"10.1016/j.fusengdes.2025.115297","url":null,"abstract":"<div><div>The technological exploitation of deuterium-tritium (DT) campaigns at JET under EUROfusion work package PrIO (Preparation of ITER Operations) aimed at taking advantage of the ITER-relevant radiation fields to improve the knowledge of nuclear technology and safety and to develop and validate nuclear codes, data and measurement techniques through dedicated experiments. Among them, the shutdown dose rate (SDDR) measurement with a dosimetry system based on ion chambers for the validation of the numerical tools for the SDDR assessment in ITER, the on-line measurement of tritium production in the mock-up of the ITER HCPB-TBM (Helium Cooled Pebble Bed - Test Blanket Module) with a diamond detector to test some measurement systems capable of working in harsh environments as TBMs, and the first-of-a-kind experiment to measure the neutron-induced activation of cooling water in a tokamak during DT operation, to validate the multi-physics methodologies for water activation assessment in ITER. Moreover, as one of the main sources of radiological hazard in ITER is due to the Activated Corrosion Products (ACPs) circulating in the cooling system, an experiment is currently under design at the ENEA 14-MeV Frascati Neutron Generator (FNG) to prove the accuracy of the reference code in this matter for ITER (i.e., OSCAR-Fusion) under fusion relevant conditions. Finally, considering the importance of radiation-induced effects on electronics in ITER and future fusion machines, it is worth mentioning the development of a modular irradiation station at ENEA named GENeuSIS (General Experimental Neutron System Irradiation Station), to be installed at FNG and which aims at reproducing specific neutron and gamma energy spectra for studying the response of electronic devices and diagnostics.</div><div>The scope of the present work is to give an overview of the mentioned activities, mostly from an experimental point of view, to describe their status, achievements, criticalities emerged and the optimization of the next steps.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115297"},"PeriodicalIF":1.9,"publicationDate":"2025-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144490408","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":"Thermo-structural assessment of the limiter inboard first wall design of the Divertor Tokamak Test facility","authors":"Riccardo De Luca ,&nbsp;Maurizio Furno Palumbo ,&nbsp;Paolo Frosi ,&nbsp;Gabriele De Sano ,&nbsp;Matteo Iafrati ,&nbsp;Gian Mario Polli ,&nbsp;Bruno Riccardi ,&nbsp;Selanna Roccella","doi":"10.1016/j.fusengdes.2025.115281","DOIUrl":"10.1016/j.fusengdes.2025.115281","url":null,"abstract":"<div><div>The Divertor Tokamak Test facility (DTT) aims at investigating integrated power exhaust solutions that can be relevant for DEMO and future power plants. Such an ambitious goal imposes several constraints on the engineering design of the actively cooled plasma-facing components (PFCs) of DTT. For instance, the First Wall (FW) must withstand thermal and electromagnetic loads that arise during both normal and off-normal operations of various plasma scenarios. In particular, the Limiter Inboard FW (LIFW), covering 50 % of the IFW, has been designed to cope with plasma limited configurations, i.e. when the plasma interacts with the solid wall. Each module consists of seven long (2.3 m) coaxial pipes made of CuCrZr alloy. Owing to the high heat loads expected, the LIFW PFCs are based on the ITER-like W-monoblock design and the plasma-facing surface, protruding radially towards the plasma with respect to the standard IFW, has a toroidal shaping that helps distribute evenly the heat load. In the present work, the technological limits of the proposed LIFW design are assessed. Based on the hydraulic conditions of the cooling water, the maximum power that can be handled by the LIFW system is evaluated under the assumption of a safety margin from the critical heat flux (CHF). Moreover, the thermo-structural behavior of a LIFW unit is simulated in ANSYS under realistic boundary conditions. In this context, a parametric distribution of the thermal load is modelled as a function of the input power and the expected spatial-temporal evolution of the plasma “footprint”. Moreover, realistic kinematic boundary conditions, representative of the pinned supports, have been included in the structural integrity assessment of the pipe, carried out according to the ITER SDC-IC design criteria (design-by-analysis approach). Preliminary results suggest that the maximum peak heat flux that can be handled by the LIFW design falls in the range 5–8 MW/m². This range is compatible with the DTT “Day0” scenario, when, due to the lesser knowledge of machine control, the most critical limiter operations may occur. Nonetheless, studies on the full power scenarios confirmed that in the ramp-up phase the maximum conductive heat load shall be lower than 1 MW/m² therefore the calculated performances can be considered adequately safe. After the fabrication of small-scale mock-ups, the lifetime of such components will be assessed experimentally, by means of cyclic thermal fatigue high heat flux tests.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115281"},"PeriodicalIF":1.9,"publicationDate":"2025-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144481644","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":"Corrigendum to “Effect of vacancy, self-interstitial atoms and transmutation rhenium on lattice thermal conductivity in tungsten” [Fusion Engineering and Design, 218 (2025), 115195]","authors":"Daigo Kanamori ,&nbsp;Keisuke Mukai ,&nbsp;Takeo Hoshi ,&nbsp;Takuya Nagasaka","doi":"10.1016/j.fusengdes.2025.115272","DOIUrl":"10.1016/j.fusengdes.2025.115272","url":null,"abstract":"","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"219 ","pages":"Article 115272"},"PeriodicalIF":1.9,"publicationDate":"2025-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144579777","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}