Fusion Engineering and Design最新文献

{"title":"Comparative analysis of the electromagnetic loads acting on the EU-DEMO divertor with alternative material options","authors":"Gennaro Di Mambro ,&nbsp;Antonio Maffucci ,&nbsp;Giuseppe Mazzone ,&nbsp;Jeong-Ha You ,&nbsp;Salvatore Ventre ,&nbsp;Fabio Villone","doi":"10.1016/j.fusengdes.2025.115024","DOIUrl":"10.1016/j.fusengdes.2025.115024","url":null,"abstract":"<div><div>This study is focused on the computation of the electromagnetic (EM) loads acting on the divertor of the EU- DEMO tokamak, related to the transients associated to plasma disruptions and to the ferromagnetic effects due to the use of materials with magnetic properties (EUROFER97). Specifically, plasma disruption events result in transient eddy and halo currents flowing into the divertor, which is then subjected to loads due to the interaction of the magnetic fields with these currents. This problem is here analyzed by using an Ansys- EMAG 3D magneto-quasi-static model simulating a toroidal sector of the tokamak that includes three divertors accurately detailed in their geometric features, and the main external components, such as blankets and vessel, described with a coarser definition. Equivalent current sources are used to impose the excitations associated to the considered conditions. The ferromagnetic effects are instead computed by using an Ansys- Maxwell magneto-static model, in which the EM loads are associated to the interaction between the magnetization of the ferromagnetic components and the known equilibrium magnetic field.</div><div>The paper carries out a comparative analysis on the EM loads associated with different design options in terms of material choices for some components: (i) vertical targets support made of AISI 316 SS, EUROFER97 or CuCrZr; (ii) pipe manifolds and supports of the cooling system made of AISI 316 SS or EUROFER97; (iii) PFC tubes and manifolds insulated or not with respect to the vertical targets support; (iv) main tubes and supports insulated or not with respect to the cassette. The impact of these choices on the estimated value of the forces and moments acting on the divertor is discussed.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115024"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808211","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":"Assessment of KSTAR Nb3Sn superconducting magnet property after long term operation since 2008","authors":"Hyun Jung Lee,&nbsp;Mu Yong Kim,&nbsp;Young Min Park,&nbsp;Giil Kwon,&nbsp;Hyun Wook Kim,&nbsp;Si-Woo Yoon","doi":"10.1016/j.fusengdes.2025.115020","DOIUrl":"10.1016/j.fusengdes.2025.115020","url":null,"abstract":"<div><div>KSTAR has been operated using Nb₃Sn superconducting magnets since 2008, which are capable of generating high magnetic fields and are ideal for high-performance plasma research. The KSTAR PF superconducting magnets have undergone repeated thermal cycling over many campaigns (16 cycles) and have been subjected to significant mechanical and thermal stress during over 30,000 plasma experiments conducted in pulse mode. To monitor any changes, a performance test is conducted on the PF1UL magnets before each campaign. These tests annually apply 4 kA and 15 kA currents, with heat generation steadily decreasing until 2024, where it has now stabilized. The KSTAR PF coils were designed to handle up to 25 kA. However, they typically operate in the range of ± 15 kA during the plasma experiments. In 2012, a 25 kA current test was conducted for the first time following the upgrade of the magnet power supply and the motor generator. Subsequently, a second 25 kA evaluation was performed in 2024 to investigate potential degradation of superconducting magnet. In this paper, we compare the thermal load on the PF1UL magnet at 25 kA charging current between 2012 and 2024 and analyze AC losses to confirm changes in its properties.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115020"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808278","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":"Integration studies of a positive neutral beam injector system into the design of a volumetric neutron source","authors":"T. Franke ,&nbsp;C. Bachmann ,&nbsp;V. Claps ,&nbsp;A. Doering ,&nbsp;T. Fellinger ,&nbsp;C. Gliss ,&nbsp;T. Haertl ,&nbsp;C. Hopf ,&nbsp;M. Kannamüller ,&nbsp;D. Leichtle ,&nbsp;R. Mozzillo ,&nbsp;J. Hun Park ,&nbsp;P. Pereslavtsev ,&nbsp;R. Riedl ,&nbsp;M. Siccinio ,&nbsp;A. Valentine","doi":"10.1016/j.fusengdes.2025.115035","DOIUrl":"10.1016/j.fusengdes.2025.115035","url":null,"abstract":"<div><div>A feasibility study regarding a volumetric neutron source (VNS) is presently conducted in the EUROfusion Consortium. The VNS uses Positive Neutral Beam Injection (P-NBI) for plasma heating, current drive and particularly to drive beam-target fusion, aiming for a high neutron production (<span><math><mo>≈</mo></math></span>0.5 MW/m² neutron wall load in the equatorial plane). P-NBI is a reliable auxiliary heating system, widely employed in plasma devices such as ASDEX Upgrade (AUG), W7-X and JT60-SA.</div><div>The paper describes the integration of the P-NBI system into the design of the VNS. This includes the neutral beam (NB) duct from the torus vacuum vessel to the NBI box with the gate valves, the space requirements of the NB injectors in the building as well as the connections for cooling water, cryo-supplies, electrical high voltage and radio frequency (RF) cable connections. Since the VNS is a nuclear machine with lifetime doses significantly exceeding those of ITER, the requirements for remote maintenance (RM) are very different from any existing NBI design. The RM concept and the RM sequences will be discussed.</div><div>Neutronics studies for the NB duct were performed and design iterations undertaken to assure that neutron heating and lifetime fluences remain below limits at the superconducting toroidal and poloidal field coils adjacent to the NB duct.</div><div>The work considers the return of experience from AUG, ITER and some other fusion experimental machines in operation, under construction or in conceptual design such as DEMO.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115035"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808215","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":"Characterization of water adsorption capacity at high pressure of the molecular sieve proposed to be used in TER RMSB","authors":"George Ana,&nbsp;George Bulubasa,&nbsp;Alina Niculescu,&nbsp;Maria Craciun,&nbsp;Ciprian Bucur,&nbsp;Iuliana Stefan","doi":"10.1016/j.fusengdes.2025.115039","DOIUrl":"10.1016/j.fusengdes.2025.115039","url":null,"abstract":"<div><div>This study investigates the effects of total pressure and superficial velocity of purge gas on the adsorption capacity of Na-Y zeolite, a candidate material for the Tritium Extraction and Recovery (TER) system in the Helium Cooled Pebble Bed (HCPB) breeding blanket. The TER system captures tritiated water vapor from helium purge gas at intended high pressure (8 MPa) to reduce pressure differentials with the cooling gas and enhance system resilience. Experimental data reveal that, while total pressure increases marginally enhance adsorption capacity, higher purge gas velocities reduce it. Adsorption isotherms were evaluated using Langmuir, Sips, and Freundlich models, with the Sips model providing the closest fit to experimental data. Mass transfer zone (MTZ) analysis indicates a stable MTZ shape along the test bed, supporting scalability of the findings for larger TER systems. Calculated utilization factors remain around 0.93–0.94 across tested pressures but decrease with increasing superficial velocity, further informing optimal operational parameters for effective adsorption and scalability in full-scale equipment.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115039"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808216","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 of three simulation tools for open-surface liquid metal magnetohydrodynamic flows in plasma-facing components using OpenFOAM","authors":"D. Suarez,&nbsp;S. Smolentsev","doi":"10.1016/j.fusengdes.2025.115008","DOIUrl":"10.1016/j.fusengdes.2025.115008","url":null,"abstract":"<div><div>The use of open-surface liquid metal (LM) flows in plasma-facing components (PFC), namely the divertor and the first wall (FW) in large tokamak nuclear fusion reactors, promises several advantages compared to solid designs, such as reduced erosion and impurity retention. The current understanding of the flow dynamics governing open-surface liquid metals for PFC applications is incomplete. In this study, we develop and test three computational models to simulate open-surface magnetohydrodynamic (MHD) flows based on the OpenFOAM computational fluid dynamics (CFD) platform. The first model is 3D and is based on the volume of fluid (VoF) method for the interface capturing and on the electric potential formulation for the MHD effects. The second model is 2D. It also uses the VOF method but relies on the integration of the MHD effect in the plane perpendicular to the magnetic field; it allows for reducing one dimension in the direction of the magnetic field. The third model is 1D, it captures the interface using a height function, which simplifies the dimension perpendicular to the substrate, and also leverages the integration of the MHD effects in the plane perpendicular to the magnetic field. Our models couple several physical phenomena and involve special numerical techniques. The three models are validated against existing experimental results for LM flows in a horizontal non-conducting chute under a co-planar magnetic field. Further simulations of the same chute with electrically conducting walls are also presented in this work, showing the capability of the 3D model to couple different regions. While the 3D model identified well the complex flow behavior and the increase in the MHD drag, the simplified 1D and 2D models failed at providing accurate estimates of flow velocity and film height. The increased drag results in the formation of an MHD-induced hydraulic jump after certain distance from the inlet nozzle. The development of these models is a necessary step towards a better understanding of LM flows, which enables a successful design of technologies capable of protecting the PFC of a fusion reactor.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115008"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808138","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":"Scoping studies on ITER tungsten first wall and borated water in Vacuum Vessel","authors":"Alberto Bittesnich ,&nbsp;Davide Laghi ,&nbsp;Marco Fabbri ,&nbsp;Alfredo Portone","doi":"10.1016/j.fusengdes.2025.115030","DOIUrl":"10.1016/j.fusengdes.2025.115030","url":null,"abstract":"<div><div>The 2023 ITER rebaselining includes the replacement of beryllium (Be) First Wall (FW) with tungsten (W). Furthermore, at the March 2023 Technical Coordination Meeting (TCM) the investigation of borated water in Vacuum Vessel (VV) – Primary Heat Transfer Systems (VV-PHTS) has been requested. In this context, scoping studies were made by using ITER C-Model, to assess the impact of these design changes on nuclear heating of different Plant Breakdown Systems (PBS). The pre-and post-processing of the calculations has been carried out using F4Enix, a new open-source Python package for Monte Carlo simulations input and output files parsing developed at Fusion for Energy (F4E). This paper wants to show and provide physical explanation of the results of the scoping studies, with particular attention to radiation shielding of superconducting magnets. The analysis evidenced that a thickness of at least 3 mm of W is needed to reach the shielding performance of 10 mm of Be, and that borated water provides relevant nuclear heating reduction mainly in the Toroidal Field Coils (TFCs), but its usability is limited by different technical reasons.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115030"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799762","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":"Comparative study of PVD techniques for tungsten deposition onto graphite substrate for fusion application","authors":"Alžběta Endrychová ,&nbsp;Slavomír Entler ,&nbsp;Vojtěch Smolík ,&nbsp;Tomáš Polcar ,&nbsp;Jan Walter ,&nbsp;Jan Podaný","doi":"10.1016/j.fusengdes.2025.115016","DOIUrl":"10.1016/j.fusengdes.2025.115016","url":null,"abstract":"<div><div>Graphite tiles are used as heat shields to protect components exposed to high-temperature plasma due to their high thermal resistance, and low atomic number, which reduces the maintenance requirements and the effect of plasma contamination. However, sputtering and hydrogen affinity are the main limitations for graphite. Tungsten has high energy threshold for sputtering and resistance to chemical erosion. Application of tungsten layer on graphite substrate may overcome graphite's limitations. This study explores the HiPIMS technique for depositing tungsten coating and tungsten coating with molybdenum interlayers onto graphite substrates. The aim was to study the influence of surface roughness on the adhesion, mechanical properties, and microstructure of coatings to optimize coated graphite for plasma-facing materials.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115016"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799760","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":"Approaching a deuterium quantitative determination methodology in tungsten matrices","authors":"G.D. Soria ,&nbsp;M. González ,&nbsp;M. Roldán ,&nbsp;G. García ,&nbsp;D. Abejón-Arribas","doi":"10.1016/j.fusengdes.2025.114992","DOIUrl":"10.1016/j.fusengdes.2025.114992","url":null,"abstract":"<div><div>In this research, a methodology for quantifying deuterium concentrations in tungsten is approached using secondary ion mass spectrometry (SIMS), a well-stablished and effective analytical technique. Tungsten standards implanted with three distinct deuterium fluences at 600 keV were prepared using a tandetron accelerator, monitoring the ion current via a Faraday cup. In order to assess the extent of damage caused by D irradiation, the microstructure and chemical composition of the W sample with the highest fluence were examined by diffraction and microscopy techniques. The precision of deuterium quantification was validated through complementary measurements performed via nuclear reaction analysis (NRA) and magnetic sector SIMS on Si testers. A calibration curve was established by correlating signals with the implanted fluences in the tungsten standards, enabling reliable quantitative deuterium detection using SIMS technique. The approach developed is critical for understanding deuterium behavior in tungsten, a key plasma-facing material in nuclear fusion reactors, where accurate quantification is essential for evaluating material performance in terms of irradiation damage and tritium inventory management.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114992"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143808210","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":"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}