Fusion Engineering and Design最新文献

{"title":"MuVacAS: Experimental setup for testing mitigation strategies against Loss of Vacuum Accidents in the IFMIF-DONES accelerator","authors":"A. Sabogal ,&nbsp;C. Torregrosa-Martín ,&nbsp;D. Rodríguez ,&nbsp;C. Caballero ,&nbsp;I. Podadera ,&nbsp;A. Ibarra ,&nbsp;D. Jimenez-Rey ,&nbsp;F. Martin-Fuertes ,&nbsp;A. Delgado ,&nbsp;U. Ruiz de Gopegui ,&nbsp;K. Garmendia","doi":"10.1016/j.fusengdes.2025.115473","DOIUrl":"10.1016/j.fusengdes.2025.115473","url":null,"abstract":"<div><div>The IFMIF-DONES facility will irradiate prospective fusion reactor materials using a linear accelerator to collide a deuteron beam with a lithium target in a liquid state. A particularity of the IFMIF-DONES accelerator is the lack of physical separation between the target vacuum chamber (TVC), where liquid Li circulates, and the vacuum chambers of the upstream accelerator components. Loss of Vacuum Accidents (LOVAs) could thus result in the potential transport of material between the TVC and the rest of the accelerator, with the risk of potential reactions with the liquid Li or the mobilisation of radioactive inventory. This paper describes the Multipurpose Vacuum Accident Scenarios (MuVacAS) experimental setup, specifically designed to validate mitigation strategies against of such accidents. MuVacAS replicates the key characteristics of the last 49 m of the IFMIF-DONES accelerator vacuum chambers. It is equipped with modules to reproduce different leak scenarios, including sudden air inrushes or the controlled injection of various gases and water. Initial experimental results from a sudden air inrush scenario are presented, showcasing the capability of MuVacAS to accurately reproduce accident conditions in vacuum systems. Furthermore, the measurement systems were reliable during the tests, ensuring the data were valid and accurate. The planned experimental campaigns aimed at reproducing a variety of LOVAs are also outlined. The data collected in these campaigns will be used to assess the effectiveness of the Fast Safety Isolation Valves (FSIVs) — the primary mitigation devices for these accidents — and to support the design of the IFMIF-DONES safety systems.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115473"},"PeriodicalIF":2.0,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267007","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":"Defect structure evolution in early-stage neutron irradiated F82H and its model alloys at 363 K and 563K","authors":"Pinyao Wang ,&nbsp;Shaosong Huang ,&nbsp;Koichi Sato ,&nbsp;Qiu Xu ,&nbsp;Toshimasa Yoshiie","doi":"10.1016/j.fusengdes.2025.115470","DOIUrl":"10.1016/j.fusengdes.2025.115470","url":null,"abstract":"<div><div>To understand the point defect processes during the incubation period of the void swelling is critical for the development of nuclear structural materials. The defect structures in the incubation period in neutron irradiated Fe and 6 model alloys of F82H, increasing the elements from Fe, were studied by positron annihilation lifetime spectroscopy and transition electron microscopy. During the irradiation at 363 K, vacancy clusters were detected in Fe and all model alloys to a dose of 0.2 dpa. With 563 K irradiation, no vacancy clusters were detected except for Fe. These results suggest the effect of Cr on model change of the long-range migration of these interstitial-type loops at high temperature. The start of steady state swelling is the saturation of the effective number of Cr atoms interacting with Fe interstitial clusters.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115470"},"PeriodicalIF":2.0,"publicationDate":"2025-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221043","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":"Simulation of neutral beam injection towards fast ion study in Thailand Tokamak-1","authors":"Apiwat Wisitsorasak ,&nbsp;Kunihiro Ogawa ,&nbsp;Siriyaporn Sangaroon ,&nbsp;Boonyarit Chatthong ,&nbsp;Akihiro Shimizu ,&nbsp;Worathat Paenthong ,&nbsp;Suebsak Suksaengpanomrung ,&nbsp;Arlee Tamman ,&nbsp;Nopporn Poolyarat ,&nbsp;Mitsutaka Isobe","doi":"10.1016/j.fusengdes.2025.115455","DOIUrl":"10.1016/j.fusengdes.2025.115455","url":null,"abstract":"<div><div>Simulations of neutral beam injection (NBI) are performed using the Monte Carlo code NUBEAM to assess NBI utilization. The study investigates various injection conditions, including beam energy and direction. Results indicate that a 20 keV beam launched in the co-current direction with a tangential radius of 0.55–0.65 m is suitable for Thailand Tokamak-1. Incorporating the toroidal field (TF) ripple, the simulations show that most beam ions escape the last closed flux surface (LCFS) in the region between two adjacent TF coils due to banana drift loss and <span><math><mrow><mo>∇</mo><mi>B</mi></mrow></math></span> drift effect. Furthermore, Alfvén eigenmodes are computed using the AE3D code, based on three-dimensional equilibria reconstructed by the VMEC code. The analysis reveals that the gap frequencies occur within the range of 130–300 kHz. The minimum beam energy required to excite the TAE modes is 2 keV at the sideband resonance condition.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115455"},"PeriodicalIF":2.0,"publicationDate":"2025-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221042","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":"Mechanical design and performance simulation of the metal foil pump for DEMO application","authors":"Changhao Wen ,&nbsp;Zhaoxi Chen ,&nbsp;Qingxi Yang ,&nbsp;Kexin Lin","doi":"10.1016/j.fusengdes.2025.115476","DOIUrl":"10.1016/j.fusengdes.2025.115476","url":null,"abstract":"<div><div>Fuel cycle research is critical to the development of sustainable fusion energy, as it enables efficient fuel utilization, tritium management, and safe, closed-loop operation of fusion reactors. To meet the fuel cycle and impurity removal requirements for subsequent fusion reactors, the Direct Internal Recycling (DIR), which employs Metal Foil Pump (MFP) technology, has been proposed. Although theoretical calculations and preliminary experiments have demonstrated the technical feasibility of the pump and its associated systems, research on the engineering application of the pump is still in its infancy. This paper theoretically investigates the hydrogen permeation through the metal foil in the MFP and analyzes the key factors influencing permeation flux. Subsequently, the pump for the future installation is designed based on actual spatial constraints and design requirement of the overall work efficiency. Finally, a brief multiphysics analysis of the designed pump is carried out to verify the performance of the designed pump. This work aims to establish foundational references for subsequent pump design iterations and to conduct early-stage research for the future practical applications of this pump.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115476"},"PeriodicalIF":2.0,"publicationDate":"2025-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220978","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 investigation of tritium release from LiAlO2 and absorption into Zr","authors":"Seiyo Kobayashi ,&nbsp;Kazunari Katayama ,&nbsp;Hiroki Isogawa ,&nbsp;Hideaki Matsuura ,&nbsp;Kenta Akashi ,&nbsp;Yuto Iinuma","doi":"10.1016/j.fusengdes.2025.115471","DOIUrl":"10.1016/j.fusengdes.2025.115471","url":null,"abstract":"<div><div>Tritium (T) production using a high-temperature gas-cooled reactor (HTGR) has been proposed to arrange initial T inventory for nuclear fusion reactors. For the establishment of this method, the behavior of T in a Li-loading rod (Li-rod), which contains LiAlO₂ and zirconium (Zr), needs to be understood. In this study, a powdered LiAlO₂ sample was heated in H₂ containing Ar gas flow to observe water formation reaction. Then, a powdered Zr sample was heated in humid Ar gas flow to investigate the temperature dependence of H₂O reduction rate on the Zr. Finally, a Zr-mixed LiAlO₂ sample, which had been neutron-irradiated in the Kyoto University Research Reactor (KUR), was preheated at 700 °C in a sealed quartz tube, and then heated in Ar gas flow to observe the release behavior of T from the sample. Released gaseous T (HT, T₂) and tritiated water vapor (HTO, T₂O) were separately collected using a pair of bubblers installed upstream and downstream of a copper oxide (CuO) bed, and the tritium concentration in each bubbler was measured using a liquid scintillation counter. Based on the results, it is suggested that LiAlO₂ should be in advance heated in reduction atmosphere to remove adsorbed water and unstable oxygen to suppress the release of T in chemical form of HTO, and Li-rod should be operated at higher temperatures over 700 °C to decrease the mass transfer resistance by an oxide layer on Zr.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115471"},"PeriodicalIF":2.0,"publicationDate":"2025-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221041","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":"Multiphysics modelling of the Activated Corrosion Products generation and transport in the WCLL PbLi loop with GETTHEM","authors":"Fabrizio Lisanti ,&nbsp;Alex Aimetta ,&nbsp;Pietro Arena ,&nbsp;Roberto Bonifetto ,&nbsp;Antonio Froio","doi":"10.1016/j.fusengdes.2025.115454","DOIUrl":"10.1016/j.fusengdes.2025.115454","url":null,"abstract":"<div><div>One of the main milestones towards the development of the EU DEMO reactor is to demonstrate the feasibility of a closed tritium fuel cycle, a key aspect for the generation of electricity from fusion energy by the middle of the century. In view of this, the design of the breeding blanket (BB) has a key role. A candidate design for the EU DEMO BB is the Water-Cooled Lithium-Lead (WCLL) concept, where eutectic lithium-lead (PbLi) is circulated in a suitable closed circuit. A key issue in the design of the PbLi circuit is the evaluation of the inventories of Activated Corrosion Products (ACPs), which are solid particles corroded from structural materials and eventually activated in the blanket, transported inside the loop within the PbLi. In recent years, a PbLi loop model has been implemented in the GETTHEM code, a system-level tool for the thermal-hydraulic modelling of BB and related subsystems. In this work, in addition to the already existing assessment of corrosion phenomena, models of different pieces of physics necessary for a comprehensive assessment of the ACP inventories are added to the PbLi loop model in GETTHEM. Specifically, these include activation and decay of the corroded species in the BB. For the latter, a sink term for the radioactive decay and a source term for the transmutation due to neutrons interaction with materials are introduced in the mass conservation equations for each ACP. To demonstrate the code capabilities, a representative test case is presented.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115454"},"PeriodicalIF":2.0,"publicationDate":"2025-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221038","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 behavior of electrical post-insulators for the MITICA Beam Source Mock-Up SOFT 2024","authors":"B. Spolaore ,&nbsp;S. Denizeau ,&nbsp;G. Chitarin ,&nbsp;M. Valente ,&nbsp;A. Rizzolo ,&nbsp;G. Berton ,&nbsp;M. Dalla Benetta ,&nbsp;N. De Rossi","doi":"10.1016/j.fusengdes.2025.115356","DOIUrl":"10.1016/j.fusengdes.2025.115356","url":null,"abstract":"<div><div>Previous R&amp;D activities conducted on the MITICA Beam Source (BS) identified the need for an experimental validation of the 1 MV High Voltage (HV) holding in vacuum of its accelerator, using a full size Mock-up that reproduces the external geometry of MITICA BS relevant for HV holding. This paper describes the analyses and tests of a critical structural component of this Mock-Up, referred to as “post-insulators”. This component is made of PolyEtherEtherKetone (PEEK), an engineering thermoplastic, and has a double-function of electrical insulator between the stages of the accelerator and of mechanical support carrying the cantilevered structure of the Mock-Up. Being critical components withstanding high loads, a Design by Experiment approach has been followed to complete the Design by Code and Design by Analysis of the PEEK insulators conducted before the construction of the experiment.</div><div>In the analyses, the stress–strain pattern of the PEEK post-insulators is simulated using two FEM models, so as to evaluate the behavior during operation and to support the interpretation of the results obtained during subsequent tests. A first model uses accurate geometry, including screws, threaded inserts and bolts connections used to fasten the PEEK post-insulator to the stainless-steel flanges. The second model is a simplified version, where insulator and flanges are connected directly with bonded contact type.</div><div>Experimental tests are performed on four post-insulator samples, using a uni-axial hydraulic test system and a particular test set-up, in order to establish the quasi-static mechanical behavior of the component under a precise combination of tensile-bending stress.</div><div>The test results are positive, however, the numerically-derived stiffness is overestimated compared to the experimental data. Such discrepancy has consequences in terms of total deformations, equivalent stress and natural frequencies of the MITICA Mock-Up BS. These differences have been quantified by calibrating the numerical models of the tensile test with the experimental results, and the Mock-Up structural simulations have been re-ran with the real insulator stiffness. Lastly, an explanation for this discrepancy is investigated by developing an accurate sub model of the M14 screw, threaded insert and PEEK bulk material, showing how modeling the screw-threaded insert connections with bonded connections overestimates the stiffness of the model compared to the real-case scenario.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115356"},"PeriodicalIF":2.0,"publicationDate":"2025-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221039","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":"Competitive effects of deuterium and hydrogen in RAFM steels under gas mixture co-exposure","authors":"Caibin Liu ,&nbsp;Fei Sun ,&nbsp;Xuechun Li ,&nbsp;Zihan Tao ,&nbsp;Haodong Liu ,&nbsp;Qiang Qi ,&nbsp;Liqun Shi ,&nbsp;Guangnan Luo ,&nbsp;Haishan Zhou","doi":"10.1016/j.fusengdes.2025.115469","DOIUrl":"10.1016/j.fusengdes.2025.115469","url":null,"abstract":"<div><div>In the deuterium (D)- tritium (T) operation phase, T retention in the structural materials of the tritium breeding blanket will raise safety concerns. These materials will face D-T mixture conditions, making it essential to understand hydrogen isotope effects on retention in RAFM steels. This work investigated D retention behavior in RAFM steels CLF-1 under 1:1 H₂+D₂ mixture gas and pure D₂ gas exposure for various duration times at 673 K. D retention was detected by thermal desorption spectrum (TDS) and elastic recoil detection analysis (ERDA) experiments. Relative to 0.05 MPa D₂, co-exposure to 0.1 MPa 1:1 H₂+D₂ (i.e., the same D₂ partial pressure) reduced low-temperature (&lt;650 K) D retention by 39 %. Meanwhile, the ERDA results showed that the maximum D concentration at the near surface (&lt;100 nm depth) was 0.16 % under the condition of 0.05 MPa pure D₂ exposure. As a comparison, the D concentration was only 0.03 % under the condition of 0.1 MPa 1:1 H₂+D₂ mixture gas exposure, indicating an 81 % reduction. The TDS and ERDA results suggested that the introduction of H₂ reduced the D retention in the RAFM steels and decreased the D concentration in the surface area. In addition, the D retention at low temperature and the surface D concentration in the RAFM steels have reached saturation within the uncertainty of the measurements when the exposure duration times of 0.1 MPa pure D₂ gas and 0.1 MPa 1:1 H₂+D₂ mixture gas exceeded 2 h.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115469"},"PeriodicalIF":2.0,"publicationDate":"2025-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220979","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":"The impact of transient heat flux on tungsten fiber-reinforced tungsten materials","authors":"Zhao Tianyu ,&nbsp;Xu Shaoqiang ,&nbsp;Du Juan ,&nbsp;Wen Pan ,&nbsp;Feng Fan ,&nbsp;Li Jialin ,&nbsp;Tang Jun ,&nbsp;Jin Fanya ,&nbsp;Zhang Kejia","doi":"10.1016/j.fusengdes.2025.115475","DOIUrl":"10.1016/j.fusengdes.2025.115475","url":null,"abstract":"<div><div>Tungsten fiber-reinforced tungsten materials (W_f/W_m), as an emerging plasma-facing material, hold promise in addressing the brittleness issues associated with traditional tungsten materials. However, the impact of transient thermal loads in nuclear fusion reactors, particularly on the tungsten fibers and interfacial structures, has been less studied, which hampers the further application and development of these materials. This paper simulates Edge Localized Modes (ELMs) events using an EMS-60 electron beam facility to subject W_f/W_m to transient thermal shock and compares the changes in fibers and interfaces before and after the transient thermal shock. The following conclusions are drawn: (1) Compared to pure tungsten materials,W_f/W_m exhibit superior performance under transient thermal loads. (2) Although in this study, the tungsten fibers and interfaces, when directly exposed to high thermal loads, can still effectively prevent crack propagation, they both experience significant performance degradation and loss of integrity. We infer that if the number of thermal shock cycles is further increased, the likelihood of failure of the fibers and interfaces may be significantly enhanced. Therefore, special attention should be paid to the failure of W_f/W_m caused by plasma flux in practical applications. (3) Tungsten fibers have an excellent inhibitory effect on cracks along the heat flux direction, and their ideal arrangement is beneath the material surface, at a position hundreds of micrometers away from the surface, where the fibers and interfacial structures can be preserved intact, effectively impeding crack propagation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115475"},"PeriodicalIF":2.0,"publicationDate":"2025-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221040","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 design and thermal analysis of high-performance titanium targets for compact D-D neutron generator","authors":"Chenglong Geng ,&nbsp;Wei Xu ,&nbsp;Zinan Feng ,&nbsp;Weihai Wang ,&nbsp;Xu Li ,&nbsp;Chenlun Liao ,&nbsp;Mengmeng Li ,&nbsp;Xiancai Meng ,&nbsp;Lizhen Liang ,&nbsp;Chundong Hu","doi":"10.1016/j.fusengdes.2025.115472","DOIUrl":"10.1016/j.fusengdes.2025.115472","url":null,"abstract":"<div><div>The titanium target is key component in accelerator-driven neutron generators. During beam injection, heat is deposited on the target surface, resulting in increasing of the surface temperature of the target. Deuterium implanted in the titanium film of the target is released significantly when the temperature exceeds 200°C. This work presents three different cooling concepts and simulations of heat load and deformation for D-D neutron generator. The heat load removal capability is improved by optimizing the structure of the coolant channel inside the target. The maximum target surface temperature was controlled at 73.7°C when the deuterium beam injection power was 600 W via a double-spiral cooling structure. It was found the thermal power transferred to the coolant increases with the coolant flow up to coolant velocity equal to 19 m/s. Moreover, the depth of the target coolant channel does not affect the heat load removal capability. Finally, a 5 × 10⁸ n/s neutron yield experiment with steady operation for 7 h was performed with the application of the double-spiral coolant structure titanium target.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115472"},"PeriodicalIF":2.0,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221037","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}