Fusion Engineering and Design最新文献

{"title":"Advanced magnetic diagnostic coils based on TPC technology for COMPASS-U and EU-DEMO","authors":"Pavel Turjanica ,&nbsp;Jan Reboun ,&nbsp;Ivan Duran ,&nbsp;Slavomir Entler ,&nbsp;Ladislav Viererbl ,&nbsp;Marek Simonovsky","doi":"10.1016/j.fusengdes.2025.115070","DOIUrl":"10.1016/j.fusengdes.2025.115070","url":null,"abstract":"<div><div>Magnetic diagnostic coils based on Thick Printed Copper (TPC) technology for COMPASS-U and EU-DEMO fusion reactors are introduced in this paper. Special emphasis is placed on the aspects of design and technology qualification. Firstly, the results of the tests carried out to qualify the TPC technology for the Tokamak environment are presented. Dedicated test samples were designed and temperature cycling irradiation and adhesion tests including electrical parameters measurements were executed. The aim of the TPC coil design was to achieve a high effective sensor area (up to 0.25 m²) and small dimensions compared to conventional coils. The coils can be composed from materials as Cu and Al₂O₃, AlN, or spinel (Al₂MgO₄) which ensures compatibility with elevated temperatures (at least 500 °C) and intense neutron radiation. Two coil variants, 1st with low winding density and 2nd with high winding density are proposed. Advanced layout of high-density coil winding achieving high suppression factor of cross-field sensitivity is presented. In parallel, the winding layout is also optimized to reduce parasitic radiation and temperature gradient induced side effects. Preliminary results of irradiation tests and temperature cycling test of TPC low winding density coil are presented.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115070"},"PeriodicalIF":1.9,"publicationDate":"2025-04-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143838251","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":"Advanced Multi-Step Brazing (AMSB) for fabrication of new type of W/stainless steel first-wall component with ODS-Cu intermediate layer","authors":"Masayuki Tokitani ,&nbsp;Toyo Yamashita ,&nbsp;Yukinori Hamaji ,&nbsp;Suguru Masuzaki ,&nbsp;Shun Shimabukuro ,&nbsp;Kaori Kono ,&nbsp;Naoaki Yoshida ,&nbsp;Makoto Oya ,&nbsp;Takumi Onchi ,&nbsp;Ryuya Ikezoe ,&nbsp;Hiroshi Idei ,&nbsp;Kazuaki Hanada","doi":"10.1016/j.fusengdes.2025.115066","DOIUrl":"10.1016/j.fusengdes.2025.115066","url":null,"abstract":"<div><div>The novel method “Advanced Multi-Step Brazing (AMSB)” has been developed to fabricate a new type of “divertor” and “first-wall” heat removal component in a fusion reactor. This study is focused on the latter component, in which a tungsten (W) sheet is jointed through AMSB to a stainless steel (SUS) substrate via an oxide dispersion strengthened copper (ODS-Cu) intermediate layer. The principle of AMSB is a repetitive application of the advanced brazing technique (ABT). The initial purpose of the ABT was to braze W to ODS-Cu (GlidCop®) with the Ni-11 %P filler material. Later, we confirmed that the ABT is able to produce a very tough GlidCop® and SUS (GlidCop®/SUS) joint. One of the major advantages of GlidCop®/SUS joints is physically strong tolerance against the repetitive brazing heat-cycle. Thus, a repetitive application of the ABT does not cause any negative effects against post-brazed GlidCop®/SUS joints, and hence AMSB can be applied for fabricating a single heat-removal component with multiple joint interfaces. A small-scale sample of the new type of first-wall component was fabricated through two-step brazing. At first, the GlidCop® plate was jointed to the SUS by the ABT, and then a thin W-sheet with the thickness of 0.254 mm was jointed to GlidCop®/SUS by the ABT. If a large area of the first-wall surface in the fusion reactor can be covered with such a thin W-sheet, the amount of hydrogen isotopes trapped on the first-wall surface could be significantly reduced, compared to other fabrication methods such as a vacuum plasma spray W (VPS-W). The VPS-W is theoretically less dense than a W-sheet and often contains pore structures, which could act as effective trapping sites for hydrogen isotopes. A small-scale sample of the new type of first-wall component with a W-sheet (W/GlidCop®/SUS) was successfully fabricated to overcome the above disadvantages. In addition, the W surface of the component showed low retention characteristics of hydrogen isotopes compared with other W surfaces, e.g., atmospheric plasma-sprayed W (APS-W).</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115066"},"PeriodicalIF":1.9,"publicationDate":"2025-04-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143834583","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":"Energy optimization of the EFCC power supply in DTT scenarios","authors":"Pasquale Zumbolo ,&nbsp;Raffaele Albanese ,&nbsp;Andrea G. Chiariello ,&nbsp;Antonio Iaiunese ,&nbsp;Alessandro Lampasi ,&nbsp;Marco Neri ,&nbsp;Raffaele Martone","doi":"10.1016/j.fusengdes.2025.115049","DOIUrl":"10.1016/j.fusengdes.2025.115049","url":null,"abstract":"<div><div>The Divertor Tokamak Test facility comprises a sophisticated system designed to counteract Error Fields. The actual field correction demand will depend on the final characteristics of the assembled machine and, during operations, on the specific scenario instant. In this work, a comprehensive approach for the optimal design of the Error Field Correction Coil currents during the scenario is presented; the model is based on a circuit model of magnetic sources, including the correction coils, as well as Poloidal Field coils, Central Solenoid, plasma current and eddy currents. The available degrees of freedom are exploited to minimize the energy demand from correction coils power supplies while ensuring effective reduction of Error Fields.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115049"},"PeriodicalIF":1.9,"publicationDate":"2025-04-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143838249","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":"Effect of divertor on tritium breeding in HELIAS","authors":"Tommi Lyytinen ,&nbsp;Antti Snicker ,&nbsp;Timo Bogaarts ,&nbsp;Felix Warmer","doi":"10.1016/j.fusengdes.2025.115000","DOIUrl":"10.1016/j.fusengdes.2025.115000","url":null,"abstract":"<div><div>This study presents tritium breeding calculations for the five-period helical-axis advanced stellarator (HELIAS) using the Serpent2 code. To address the absence of a detailed divertor design, a simplified divertor model with a layered homogenized structure was integrated into the parametric geometry of the HELIAS. The impact of the divertor on the tritium breeding ratio (TBR) was assessed through a parametric study, varying divertor configurations, surface area, materials for both the divertor and breeding blanket, and the thickness of the breeding zone behind the divertor. Calculated TBR losses relative to full blanket coverage ranged from 4 % to 35 %, depending on the specified input parameters. The primary factors influencing TBR loss were the fraction of surface area occupied by the divertor and the volume of the breeding zone displaced by it. A linear relationship was observed between TBR loss and both the divertor area fraction and the reduction in breeding zone volume. Furthermore, water-cooled divertor configurations showed a more negative impact on TBR compared to helium-cooled, with effects comparable in significance to variations in area fraction. The difference in TBR loss between purely TBR-optimized divertor configuration and Wendelstein 7-X-like top-bottom configuration was less than 2 percentage points.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115000"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828891","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":"Analyses of vacuum vessel pressure suppression system during divertor LOCA for Helium cooled pebble bed DEMO","authors":"G. Mazzini ,&nbsp;M. D'Onorio ,&nbsp;D.N. Dongiovanni ,&nbsp;J. Syblik","doi":"10.1016/j.fusengdes.2025.115043","DOIUrl":"10.1016/j.fusengdes.2025.115043","url":null,"abstract":"<div><div>An important aspect under investigation for several years in the EU-DEMO reactor is the mitigation of hydrogen concentration - mainly tritium and deuterium - in the Vacuum Vessel (VV) and the surrounding volumes. This accumulation poses risks, including potential explosion hazards, which could damage confinement barriers. One of the possible solutions, applicable to both water and helium-cooled blanket concepts, involves the use of Passive Autocatalytic Recombines (PAR). These devices, representing a cross-cutting technology between fission and fusion facilities, are being considered for integration into the Vacuum Vessel Pressure Suppression System (VVPSS) and related systems.</div><div>This paper focuses on the hydrogen transport analysis, specifically the inVV tritium and deuterium inventory mobilization towards the VVPSS designed for - Helium-Cooled Pebble Bed (HCPB) concept for which a model was developed by means of MELCOR 1.8.6 for fusion applications. Previous work on the HCPB Loss of Coolant Accident (LOCA) analyses suggests that inert helium resulting from Helium coolant blowdown can significantly mitigate the risk of hydrogen deflagration. The paper explores a new scenario, the Divertor in-vessel LOCA, which may lead to increased hydrogen mass due to the water-metal reaction. The paper discusses the integration of PAR technology within the operational range of fusion devices, specifically focusing on its potential to recombine hydrogen within the Expansion Volumes.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115043"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143829810","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":"Logical approach to tritium vacuum pump selection for fusion applications","authors":"Lucas M. Angelette,&nbsp;James E. Klein","doi":"10.1016/j.fusengdes.2025.114987","DOIUrl":"10.1016/j.fusengdes.2025.114987","url":null,"abstract":"<div><div>Vacuum pumps are a critical component of fusion reactors. However, there are dramatically different vacuum pumping needs for different systems in the fusion fuel cycle. There have been limited studies on the applicability of off-the-shelf pumps for use in tritium service, and there is no clear focus for directing pump manufacturers to better meet the needs for fusion applications. In this paper, the advantages and disadvantages of various pumping technologies are discussed along with their ideal point of use.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 114987"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143834582","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-mechanical analysis of plasma grid for 120 keV/50 A accelerator in EAST neutral beam injector","authors":"Fang Wang ,&nbsp;Yuanlai Xie ,&nbsp;Yahong Xie ,&nbsp;Yuming Gu ,&nbsp;Qianxu Wang ,&nbsp;Longbin Liu ,&nbsp;Yang Zhu ,&nbsp;Huihui Hong ,&nbsp;Bin Li ,&nbsp;Jianglong Wei ,&nbsp;Kun Tian","doi":"10.1016/j.fusengdes.2025.115047","DOIUrl":"10.1016/j.fusengdes.2025.115047","url":null,"abstract":"<div><div>High-current ion sources are critical components in neutral beam injection (NBI) systems for magnetic confinement fusion, with their performance directly determining the efficiency and reliability of the entire NBI system. With increasing demands for neutral beam injection heating in magnetic confinement fusion research, developing ion sources with higher beam energy and power has become a primary research focus. In response to new design requirements for 120 keV/50 A ion sources, this study investigates the thermodynamic performance of a newly designed plasma grid for a 120 keV/50 A triode accelerator under long-pulse high-power operating conditions. A thermal-stress coupling analysis model was established using numerical methods that coupled FLUENT with finite element software. Through systematic analysis, molybdenum was identified as the optimal material choice, and the minimum required cooling water mass flow rate for safe operation under long-pulse high-power conditions was determined to be 0.426 kg/s. In-depth analysis under the most severe operating conditions demonstrated that the maximum Von-Mises equivalent stress remains below the material yield strength. The reliability of the numerical model was validated through experimental results from the EAST-NBI comprehensive test platform, providing crucial evidence for safe operation under long-pulse high-power conditions. This work presents significant value for advancing the development of high-performance neutral beam injection systems.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115047"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828893","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":"Influence of Li on the behavior of H and point defects in vanadium alloys","authors":"Shuxian Sun ,&nbsp;Mingliang Wei ,&nbsp;Pengbo Zhang ,&nbsp;Yichao Wang ,&nbsp;Yaxia Wei ,&nbsp;Pengfei Zheng","doi":"10.1016/j.fusengdes.2025.115079","DOIUrl":"10.1016/j.fusengdes.2025.115079","url":null,"abstract":"<div><div>Vanadium alloys are promising structural materials for first wall and blanket applications in advanced fusion reactors. Hydrogen (H) retention strongly affects their stability during operation. Through first-principles calculations and empirical formulas, we investigated the influence of lithium (Li) on the behavior of H and point defects in vanadium alloys. The results showed that Li increases H retention and act as a stabilizing effect on H. The binding of Li-vacancy enhances vacancy-trapping ability for H impurities and the stability of H-vacancy complexes, while Li-vacancy still stably trap six H. Kinetically, H atom migrating to Li exhibits a lower barrier (0.058 eV), in turn it will overcome a high barrier (0.22 eV), indicating Li hindering H mobility. Moreover, the predictions of empirical formulas presented that Li reduces the effective diffusivity of H but increases the permeability of H in vanadium alloy. On the other hand, the Li-vacancy interaction is a weak attraction, kinetically Li decreases migration barrier of vacancy near Li by 0.34 eV, promoting vacancy diffusion in 〈111〉 direction. Li atoms can form Li-vanadium mixed interstitials and reduce the rotation barriers of 〈111〉 _{mixed V-Li}, slowing down self-interstitial diffusion. These findings deep understanding for the synergetic effects of Li with H impurities and point defects in vanadium alloys.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115079"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828894","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":"Kinetics of gas release from titanium beryllide irradiated to different neutron fluences","authors":"A. Shaimerdenov ,&nbsp;A. Akhanov ,&nbsp;S. Askerbekov ,&nbsp;T. Kulsartov ,&nbsp;M. Aitkulov ,&nbsp;D. Sairanbayev ,&nbsp;A. Nessipbay ,&nbsp;Zh. Bugybay ,&nbsp;I. Kenzhina ,&nbsp;B. Shakirov ,&nbsp;M. Kylyshkanov ,&nbsp;M. Podoinikov","doi":"10.1016/j.fusengdes.2025.115069","DOIUrl":"10.1016/j.fusengdes.2025.115069","url":null,"abstract":"<div><div>The binary intermetallic compound of beryllium and titanium (Be₁₂Ti) is an advanced material for DEMO-type fusion reactors as a neutron multiplier in breeder blankets. Titanium beryllide has a high melting point (∼1873 K), high oxidation resistance and low swelling, which makes it attractive for use in fission and fusion reactors. The study of its behavior in a radiation field is of utmost importance from the point of view of practical application of titanium beryllide in nuclear installations.</div><div>Based on this, at the Institute of Nuclear Physics (Kazakhstan), R&amp;D is underway to characterization of industrial fabricated titanium beryllide (Be₁₂Ti) at JSC “Ulba Metallurgical Plant” (Kazakhstan). Titanium beryllide samples were irradiated in the WWR-K reactor core to accumulate two integral neutron fluences of 1.6 × 10²⁵ m^-2 and 2.3 × 10²⁵ m^-2, in an inert environment and at low temperature to minimize gas migration from the samples.</div><div>The present work presents the results of post reactor studies to determine the amount of generated tritium and helium in irradiated Be₁₂Ti samples. The studies were carried out by thermodesorption spectrometry (TDS) method with mass-spectra control of gas composition. Mass spectra of gas extraction from titanium beryllide irradiated to different neutron fluences and heated up to 1173–1223 K at different heating rates (5, 10, 20 K/min) are presented. The kinetic parameters of tritium desorption from Be₁₂Ti were determined, and the activation energy was estimated for each fluence achieved.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115069"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828892","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":"80keV 1 MW NBI on COMPASS tokamak: the results and operational experience","authors":"I. Mysiura ,&nbsp;K. Bogar ,&nbsp;O. Ficker ,&nbsp;M. Komm ,&nbsp;F. Jaulmes ,&nbsp;J. Varju ,&nbsp;S. Fukova ,&nbsp;P. Vondracek ,&nbsp;M. Hron ,&nbsp;R. Panek ,&nbsp;COMPASS Team","doi":"10.1016/j.fusengdes.2025.115078","DOIUrl":"10.1016/j.fusengdes.2025.115078","url":null,"abstract":"<div><div>Prior to the final experimental campaign of the COMPASS tokamak, its heating system was upgraded with the 1 MW, 80 keV Neutral Beam Injection (NBI) unit, complementing the existing two 40 keV, 300 kW NBIs. This upgrade significantly improved tokamak performance, achieving a central electron temperature of 2 keV, a 30 % increase in plasma stored energy, an eightfold rise in neutron yield, and enabling H-mode operation in a reversed toroidal magnetic field.</div><div>Experimental tests confirmed that the system met its technical specifications, delivering 1 MW of neutral beam power with a beam divergence of &lt;12 mrad, thereby validating its suitability for COMPASS-U. However, the integration process revealed technical and design challenges, which provided valuable insights for mitigation strategies and future improvements. The COMPASS-U project plans to incorporate up to six similar injectors, making the operational experience gained from this upgrade critical for future tokamak operations.</div><div>Furthermore, the data collected during the campaign highlighted limitations in COMPASS's auxiliary systems, including water cooling performance, grounding schemes, and other infrastructure requirements. These findings will inform the optimization of the future tokamak and NBI's systems and ensure robust performance in next-generation tokamak experiments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115078"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828895","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}