Pub Date : 2026-04-01Epub Date: 2026-01-21DOI: 10.1016/j.fusengdes.2026.115631
Jianguo Ma , Zhiyong Wang , Tao Zhu , Zhihong Liu , Wangqi Shi , Huapeng Wu , Haiying Xu , Weiping Fang , Yudong Su , Jiefeng Wu
As the preferred material for plasma-facing components in future fusion test reactors, tungsten plays a critical role in ensuring the safe and stable operation of fusion reactors on the first wall of blankets and divertor targets. This paper aims to explore advanced manufacturing methods for pure tungsten and analyze the feasibility of applying additive manufacturing technology in nuclear fusion. Pure tungsten components were fabricated using powder bed fusion electron beam (PBF-EB), followed by annealing heat treatment in this work. The evolution of microstructure and mechanical properties at different annealing temperatures was investigated. Results revealed a distinct polyhedral equiaxed grain structure, with average grain size initially decreasing and then increasing as annealing temperature rose. Optimal performance was achieved at 1100 °C, with a density of 99.5%, Vickers hardness of 406 HV0.3, and compressive strength of 1961 MPa. Compared to untreated specimens, these properties showed substantial improvement. The findings provide guidance for developing properties of other refractory materials and improve the application of additive manufacturing in plasma-faced material fabrication.
{"title":"Effect of annealing on microstructure and mechanical properties of tungsten fabricated via Powder Bed Fusion Electron Beam (PBF-EB)","authors":"Jianguo Ma , Zhiyong Wang , Tao Zhu , Zhihong Liu , Wangqi Shi , Huapeng Wu , Haiying Xu , Weiping Fang , Yudong Su , Jiefeng Wu","doi":"10.1016/j.fusengdes.2026.115631","DOIUrl":"10.1016/j.fusengdes.2026.115631","url":null,"abstract":"<div><div>As the preferred material for plasma-facing components in future fusion test reactors, tungsten plays a critical role in ensuring the safe and stable operation of fusion reactors on the first wall of blankets and divertor targets. This paper aims to explore advanced manufacturing methods for pure tungsten and analyze the feasibility of applying additive manufacturing technology in nuclear fusion. Pure tungsten components were fabricated using powder bed fusion electron beam (PBF-EB), followed by annealing heat treatment in this work. The evolution of microstructure and mechanical properties at different annealing temperatures was investigated. Results revealed a distinct polyhedral equiaxed grain structure, with average grain size initially decreasing and then increasing as annealing temperature rose. Optimal performance was achieved at 1100 °C, with a density of 99.5%, Vickers hardness of 406 HV<sub>0.3</sub>, and compressive strength of 1961 MPa. Compared to untreated specimens, these properties showed substantial improvement. The findings provide guidance for developing properties of other refractory materials and improve the application of additive manufacturing in plasma-faced material fabrication.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115631"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146025281","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}
Pub Date : 2026-04-01Epub Date: 2026-01-20DOI: 10.1016/j.fusengdes.2026.115636
Yifan Zhang , Kecheng Jiang , Lei Chen , Xuebin Ma , Songlin Liu
To address the risk of flow non-uniformity in the Water-Cooled Ceramic Breeder (WCCB) blanket of the China Fusion Engineering Test Reactor (CFETR), three acrylic-based test sections were developed for flow visualization experiments: (1) 1:5 scaled outboard segment model featuring hydraulically equivalent pipes regulated by pinch valves, (2) first wall (FW) model with 39 U-shaped cooling channels, and (3) breeding zone (BZ) model comprising four groups of 29 cooling tubes each. A key contribution of this work is the development of the scaling strategy for outboard segment test section based on equal Euler number (Eu) and cooling water velocity matching, resolving scaling conflicts between the high-pressure/high-temperature prototype and the ambient-condition test section. Computational Fluid Dynamics (CFD) optimization showed that the maximum deviation in blanket module between the 1:5 scaled outboard segment, and the full-scale prototype is 6.1%. For FW test section, geometric optimizations—including inlet pipe downsizing and manifold wall thinning—reduced the maximum deviation in channel flow distribution to 0.51%, while the cooling water streamlines in the manifold closely matched those of the prototype. Mass flow rates in the cooling channels are measured using both Doppler-based ultrasonic flowmeters and Particle Image Velocimetry (PIV), enabling cross-validation and detailed characterization of the internal flow field. This test section design provides high-fidelity experimental support for the hydraulic optimization of the CFETR WCCB blanket.
{"title":"Development and optimization of flow distribution test sections for CFETR WCCB blanket","authors":"Yifan Zhang , Kecheng Jiang , Lei Chen , Xuebin Ma , Songlin Liu","doi":"10.1016/j.fusengdes.2026.115636","DOIUrl":"10.1016/j.fusengdes.2026.115636","url":null,"abstract":"<div><div>To address the risk of flow non-uniformity in the Water-Cooled Ceramic Breeder (WCCB) blanket of the China Fusion Engineering Test Reactor (CFETR), three acrylic-based test sections were developed for flow visualization experiments: (1) 1:5 scaled outboard segment model featuring hydraulically equivalent pipes regulated by pinch valves, (2) first wall (FW) model with 39 U-shaped cooling channels, and (3) breeding zone (BZ) model comprising four groups of 29 cooling tubes each. A key contribution of this work is the development of the scaling strategy for outboard segment test section based on equal Euler number (<em>Eu</em>) and cooling water velocity matching, resolving scaling conflicts between the high-pressure/high-temperature prototype and the ambient-condition test section. Computational Fluid Dynamics (CFD) optimization showed that the maximum deviation in blanket module between the 1:5 scaled outboard segment, and the full-scale prototype is 6.1%. For FW test section, geometric optimizations—including inlet pipe downsizing and manifold wall thinning—reduced the maximum deviation in channel flow distribution to 0.51%, while the cooling water streamlines in the manifold closely matched those of the prototype. Mass flow rates in the cooling channels are measured using both Doppler-based ultrasonic flowmeters and Particle Image Velocimetry (PIV), enabling cross-validation and detailed characterization of the internal flow field. This test section design provides high-fidelity experimental support for the hydraulic optimization of the CFETR WCCB blanket.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115636"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146025232","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}
Pub Date : 2026-04-01Epub Date: 2026-01-21DOI: 10.1016/j.fusengdes.2026.115633
C.H. Wang , F. Zhao , F.H. Xu , S.P. Xiong , M. Yang , W.S. Huang
China Low Activation Martensitic (CLAM) steel serves as a cladding material for thermonuclear fusion reactors. To guarantee its performance in high-temperature irradiated environments, improving the stability of precipitated phases is critical. This investigation utilized pre-precipitation thermomechanical treatment to control the precipitation site and density of the MX precipitation phase, thereby alleviating the destabilization of the precipitated phases in CLAM steel during irradiation. Heat-treated CLAM samples were subjected to Fe²⁺ ion irradiation at 450 °C, achieving fluences of 5 dpa and 15 dpa. The experimental results indicate that prior to irradiation, compared with the normalization + tempering treatment, the pre-precipitation thermomechanical treatment + tempering process resulted in refined martensitic lath structures, increased dislocation density, and preferential precipitation of the MX precipitation phase at the grain boundaries in the CLAM steel, accompanied by a reduced precipitate size and increased phase density. Post-irradiation, both lath structures and precipitates experienced coarsening; radiation-induced amorphization was observed at the M23C6 phase boundaries, whereas the MX phase retained excellent crystallinity. This study revealed that high-density, nanoscale MX phases precipitated at martensitic lath interfaces via the pre-precipitation thermomechanical treatment effectively pinned dislocations and impeded lath coarsening during irradiation. Concurrently, stable MX phases constrained partial amorphization and coarsening of adjacent M23C6 phases. These microstructural modifications enhance the irradiation-induced microstructural stability of CLAM steel, offering insights for optimizing nuclear structural materials.
{"title":"Effect of pre-precipitation thermomechanical treatment on the phase stability of CLAM steel after Fe2+ ion irradiation","authors":"C.H. Wang , F. Zhao , F.H. Xu , S.P. Xiong , M. Yang , W.S. Huang","doi":"10.1016/j.fusengdes.2026.115633","DOIUrl":"10.1016/j.fusengdes.2026.115633","url":null,"abstract":"<div><div>China Low Activation Martensitic (CLAM) steel serves as a cladding material for thermonuclear fusion reactors. To guarantee its performance in high-temperature irradiated environments, improving the stability of precipitated phases is critical. This investigation utilized pre-precipitation thermomechanical treatment to control the precipitation site and density of the MX precipitation phase, thereby alleviating the destabilization of the precipitated phases in CLAM steel during irradiation. Heat-treated CLAM samples were subjected to Fe²⁺ ion irradiation at 450 °C, achieving fluences of 5 dpa and 15 dpa. The experimental results indicate that prior to irradiation, compared with the normalization + tempering treatment, the pre-precipitation thermomechanical treatment + tempering process resulted in refined martensitic lath structures, increased dislocation density, and preferential precipitation of the MX precipitation phase at the grain boundaries in the CLAM steel, accompanied by a reduced precipitate size and increased phase density. Post-irradiation, both lath structures and precipitates experienced coarsening; radiation-induced amorphization was observed at the M<sub>23</sub>C<sub>6</sub> phase boundaries, whereas the MX phase retained excellent crystallinity. This study revealed that high-density, nanoscale MX phases precipitated at martensitic lath interfaces via the pre-precipitation thermomechanical treatment effectively pinned dislocations and impeded lath coarsening during irradiation. Concurrently, stable MX phases constrained partial amorphization and coarsening of adjacent M<sub>23</sub>C<sub>6</sub> phases. These microstructural modifications enhance the irradiation-induced microstructural stability of CLAM steel, offering insights for optimizing nuclear structural materials.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115633"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146025234","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}
Pub Date : 2026-04-01Epub Date: 2026-02-09DOI: 10.1016/j.fusengdes.2026.115660
L. van Ham , S.A. Lazerson , B.C. Hamstra , S.A. Bozhenkov , P. McNeely , N. Rust , D. Hartmann , W7-X Team
A new comprehensive model for calculating ion trajectories inside the neutral beam injection (NBI) system of Wendelstein 7-X (W7-X) is presented. The model consists of two parts. First, the magnetic materials code MUMAT has been developed to calculate the magnetic response of ferritic materials (including NBI shielding) to the magnetic field produced by the main W7-X coil system. This code has been verified through application to a scenario with a known magnetic field. Second, the Monte Carlo particle following code BEAMS3D follows particles through the resulting magnetic field and estimates heat loads on NBI components. MUMAT calculations of the magnetic field inside the NBI system predict significant fields (10 Gauss) inside the NBI neutralizer, indicating that stray magnetic fields penetrate the NBI system. Subsequent BEAMS3D simulations predict that heat loads on NBI components shift vertically due to this neutralizer field. These shifts agree qualitatively with experimental observations, and a quantitative comparison with infrared imaging and calorimetry is planned.
{"title":"MUMAT: A magnetostatics model for predicting stray magnetic fields in the Wendelstein 7-X neutral beam systems","authors":"L. van Ham , S.A. Lazerson , B.C. Hamstra , S.A. Bozhenkov , P. McNeely , N. Rust , D. Hartmann , W7-X Team","doi":"10.1016/j.fusengdes.2026.115660","DOIUrl":"10.1016/j.fusengdes.2026.115660","url":null,"abstract":"<div><div>A new comprehensive model for calculating ion trajectories inside the neutral beam injection (NBI) system of Wendelstein 7-X (W7-X) is presented. The model consists of two parts. First, the magnetic materials code MUMAT has been developed to calculate the magnetic response of ferritic materials (including NBI shielding) to the magnetic field produced by the main W7-X coil system. This code has been verified through application to a scenario with a known magnetic field. Second, the Monte Carlo particle following code BEAMS3D follows particles through the resulting magnetic field and estimates heat loads on NBI components. MUMAT calculations of the magnetic field inside the NBI system predict significant fields (<span><math><mo>≥</mo></math></span>10 Gauss) inside the NBI neutralizer, indicating that stray magnetic fields penetrate the NBI system. Subsequent BEAMS3D simulations predict that heat loads on NBI components shift vertically due to this neutralizer field. These shifts agree qualitatively with experimental observations, and a quantitative comparison with infrared imaging and calorimetry is planned.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115660"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146173793","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}
Pub Date : 2026-04-01Epub Date: 2026-01-23DOI: 10.1016/j.fusengdes.2026.115635
Chi Lei , Zhoujun Yang , Zhifeng Cheng , Nengchao Wang , Zezhi Yu , Zijian Xuan , Yan Guo , Siyu Zhu , J-TEXT Team
For magnetic confinement fusion, the measurement of hydrogen isotope ratios is of critical importance. It not only reflects the fuel ratio in future fusion reactors but also provides a quantitative analysis for investigating isotope effects on plasma confinement. In this paper, a spectral diagnostic system for measuring the ratio of hydrogen (H) and deuterium (D) at plasma edge has been developed on J-TEXT tokamak, based on the spectra of hydrogen isotope Balmer-alpha line radiation. To meet the requirement of spectra measurement for Hα (656.28 nm), Dα (656.10 nm) and Tα (656.04 nm), a high-resolution spectroscope with wavelength resolution of 0.0073 nm/pixel and time resolution of 5 ms is adopted for the diagnostic. An analysis module has been developed by employing a multi-parameter spectral shape-fitting algorithm and considering the Zeeman effect and Doppler broadening. The diagnostic has been applied in experiments, in which the H-D mixing ratio is affected by the wall conditions. The experimental results confirmed the excellent measurement ability of the high-resolution spectral diagnostic of the edge hydrogen isotope ratio developed for J-TEXT, which will provide the necessary H-D concentration information for the subsequent isotope experiments.
{"title":"Development of spectral diagnostic for edge hydrogen isotope ratio on J-TEXT","authors":"Chi Lei , Zhoujun Yang , Zhifeng Cheng , Nengchao Wang , Zezhi Yu , Zijian Xuan , Yan Guo , Siyu Zhu , J-TEXT Team","doi":"10.1016/j.fusengdes.2026.115635","DOIUrl":"10.1016/j.fusengdes.2026.115635","url":null,"abstract":"<div><div>For magnetic confinement fusion, the measurement of hydrogen isotope ratios is of critical importance. It not only reflects the fuel ratio in future fusion reactors but also provides a quantitative analysis for investigating isotope effects on plasma confinement. In this paper, a spectral diagnostic system for measuring the ratio of hydrogen (H) and deuterium (D) at plasma edge has been developed on J-TEXT tokamak, based on the spectra of hydrogen isotope Balmer-alpha line radiation. To meet the requirement of spectra measurement for Hα (656.28 nm), Dα (656.10 nm) and Tα (656.04 nm), a high-resolution spectroscope with wavelength resolution of 0.0073 nm/pixel and time resolution of 5 ms is adopted for the diagnostic. An analysis module has been developed by employing a multi-parameter spectral shape-fitting algorithm and considering the Zeeman effect and Doppler broadening. The diagnostic has been applied in experiments, in which the H-D mixing ratio is affected by the wall conditions. The experimental results confirmed the excellent measurement ability of the high-resolution spectral diagnostic of the edge hydrogen isotope ratio developed for J-TEXT, which will provide the necessary H-D concentration information for the subsequent isotope experiments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115635"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146025385","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}
Pub Date : 2026-04-01Epub Date: 2026-02-12DOI: 10.1016/j.fusengdes.2026.115661
Ayobami Daniel Daramola
The thermo-mechanical integrity of tungsten monoblock divertor targets limits high heat-flux operation in present and future fusion devices. A coupled thermo-mechanical model is presented to examine hot-spot formation, yield exceedance, and damage accumulation in water-cooled tungsten monoblocks under steady and cyclic heat loading. The model accounts for heat conduction, elastoplastic stress response, and geometry-induced mechanical constraint, with interfacial heat transfer represented through an effective contact conductance that can degrade with thermo-mechanical loading.
Baseline calculations reveal a clear separation between thermal and mechanical hot-spots. Peak temperatures are controlled primarily by heat-extraction pathways and thermal resistance, whereas the highest failure ratios and plastic strains develop near material interfaces in mechanically constrained regions above the cooling channel. Parametric variations show pronounced trade-offs between thermal and mechanical margins: geometric changes that reduce peak temperature can increase yield exceedance, while configurations with higher temperatures may exhibit reduced mechanical constraint. Sensitivity and uncertainty analyses indicate that variability in peak temperature and failure ratio is dominated by a small number of geometric parameters, particularly those governing coolant pipe position and interlayer geometry.
Under cyclic loading, plastic strain progressively localises near the cooling channel and accumulates through ratcheting. In the coupled model, this accumulation degrades interfacial heat transfer, leading to increased thermal resistance and amplification of thermo-mechanical hot-spots. While the magnitude of these effects depends on phenomenological degradation assumptions, the results demonstrate that temperature-based metrics alone do not capture structural risk. Mechanical constraint and interface behaviour play a central role in governing damage localisation and divertor performance.
{"title":"Thermo-mechanical hot-spot formation and damage accumulation in tungsten monoblock divertor targets","authors":"Ayobami Daniel Daramola","doi":"10.1016/j.fusengdes.2026.115661","DOIUrl":"10.1016/j.fusengdes.2026.115661","url":null,"abstract":"<div><div>The thermo-mechanical integrity of tungsten monoblock divertor targets limits high heat-flux operation in present and future fusion devices. A coupled thermo-mechanical model is presented to examine hot-spot formation, yield exceedance, and damage accumulation in water-cooled tungsten monoblocks under steady and cyclic heat loading. The model accounts for heat conduction, elastoplastic stress response, and geometry-induced mechanical constraint, with interfacial heat transfer represented through an effective contact conductance that can degrade with thermo-mechanical loading.</div><div>Baseline calculations reveal a clear separation between thermal and mechanical hot-spots. Peak temperatures are controlled primarily by heat-extraction pathways and thermal resistance, whereas the highest failure ratios and plastic strains develop near material interfaces in mechanically constrained regions above the cooling channel. Parametric variations show pronounced trade-offs between thermal and mechanical margins: geometric changes that reduce peak temperature can increase yield exceedance, while configurations with higher temperatures may exhibit reduced mechanical constraint. Sensitivity and uncertainty analyses indicate that variability in peak temperature and failure ratio is dominated by a small number of geometric parameters, particularly those governing coolant pipe position and interlayer geometry.</div><div>Under cyclic loading, plastic strain progressively localises near the cooling channel and accumulates through ratcheting. In the coupled model, this accumulation degrades interfacial heat transfer, leading to increased thermal resistance and amplification of thermo-mechanical hot-spots. While the magnitude of these effects depends on phenomenological degradation assumptions, the results demonstrate that temperature-based metrics alone do not capture structural risk. Mechanical constraint and interface behaviour play a central role in governing damage localisation and divertor performance.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115661"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146173747","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}
Pub Date : 2026-04-01Epub Date: 2026-02-02DOI: 10.1016/j.fusengdes.2026.115655
Yang Yang , Yu Zhang , Liansheng Du , Xuwei Zhu , Junfei Pang , Yuxiang Tang , Yong Cheng
The double door system is a critical component of the remote maintenance cask system for fusion reactors, enabling the safe transfer of activated in-vessel components between the vacuum vessel and the hot cell. This paper presents the design, development, and experimental validation of a novel double door system. A full scale, fully functional double door prototype was constructed, featuring key design elements such as a rectangular door profile, irregular cross-section sealing ring, a purely mechanical port door and cask door coupling and locking mechanism, and an integrated door tilting system. The double door prototype was used to execute a comprehensive test campaign. This included validation of the complete motion sequence — docking, door unlocking, two door coupling, tilting, and reversal — and leak tightness testing according to ISO 10648-2 standards for all relevant sealed chambers. Experimental results confirmed reliable mechanical operation and demonstrated that all tested chambers met the Class 2 and Class 3 leak tightness requirement. The successful validation of both functional and sealing performance confirms the design’s effectiveness and provides a valuable reference for the engineering of transfer cask system in future fusion power plants.
{"title":"Design and verification of the double door system for fusion reactor remote maintenance casks","authors":"Yang Yang , Yu Zhang , Liansheng Du , Xuwei Zhu , Junfei Pang , Yuxiang Tang , Yong Cheng","doi":"10.1016/j.fusengdes.2026.115655","DOIUrl":"10.1016/j.fusengdes.2026.115655","url":null,"abstract":"<div><div>The double door system is a critical component of the remote maintenance cask system for fusion reactors, enabling the safe transfer of activated in-vessel components between the vacuum vessel and the hot cell. This paper presents the design, development, and experimental validation of a novel double door system. A full scale, fully functional double door prototype was constructed, featuring key design elements such as a rectangular door profile, irregular cross-section sealing ring, a purely mechanical port door and cask door coupling and locking mechanism, and an integrated door tilting system. The double door prototype was used to execute a comprehensive test campaign. This included validation of the complete motion sequence — docking, door unlocking, two door coupling, tilting, and reversal — and leak tightness testing according to ISO 10648-2 standards for all relevant sealed chambers. Experimental results confirmed reliable mechanical operation and demonstrated that all tested chambers met the Class 2 and Class 3 leak tightness requirement. The successful validation of both functional and sealing performance confirms the design’s effectiveness and provides a valuable reference for the engineering of transfer cask system in future fusion power plants.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115655"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146173789","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}
Pub Date : 2026-04-01Epub Date: 2026-02-11DOI: 10.1016/j.fusengdes.2026.115662
Jie Liu , Xingfu Ye , Fengchao Zhao , Zhe Chen , Shen Qu
The Second-generation helium (HEL-2) cooling verification loop serves as the prototype for the Helium Cooled Ceramic Breeder Test Blanket System (HCCB TBS) of the International Thermonuclear Experimental Reactor (ITER) project. This study focuses on optimizing the insulation design of the high-pressure tank of HEL-2 to ensure structural integrity under extreme fire and seismic conditions. A transient heat conduction analysis was conducted to evaluate the thermal resistance performance of two insulating materials under extreme fire conditions. Integrating considerations of room space occupancy, cost, and installation complexity, a composite insulation layer solution was proposed. Furthermore, the impact of the anchorage structure's seismic resilience was assessed using the missing mass method combined with modal analysis. Incorporating thermal resistance performance, economic considerations, an optimal composite insulation layer scheme was identified.
{"title":"Research on the usage of insulation layer under extreme fire and seismic conditions based on transient heat conduction and missing mass methods","authors":"Jie Liu , Xingfu Ye , Fengchao Zhao , Zhe Chen , Shen Qu","doi":"10.1016/j.fusengdes.2026.115662","DOIUrl":"10.1016/j.fusengdes.2026.115662","url":null,"abstract":"<div><div>The Second-generation helium (HEL-2) cooling verification loop serves as the prototype for the Helium Cooled Ceramic Breeder Test Blanket System (HCCB TBS) of the International Thermonuclear Experimental Reactor (ITER) project. This study focuses on optimizing the insulation design of the high-pressure tank of HEL-2 to ensure structural integrity under extreme fire and seismic conditions. A transient heat conduction analysis was conducted to evaluate the thermal resistance performance of two insulating materials under extreme fire conditions. Integrating considerations of room space occupancy, cost, and installation complexity, a composite insulation layer solution was proposed. Furthermore, the impact of the anchorage structure's seismic resilience was assessed using the missing mass method combined with modal analysis. Incorporating thermal resistance performance, economic considerations, an optimal composite insulation layer scheme was identified.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115662"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146173796","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}
Pub Date : 2026-04-01Epub Date: 2026-01-14DOI: 10.1016/j.fusengdes.2025.115610
Giacomo Cavuoti , Francesca Cau , José Lorenzo , Alfredo Portone
The aim of this paper is to present a fast method capable of computing thermo-hydraulic transients in solid components that are cooled (or heated) by incompressible forced flow with or without external heating sources. By coupling the heat conduction equation in the solid volume to the heat transfer to the forced flow we derive a linearized mapping between the vector of input (control) quantities u(t) such as mass flow and inlet temperature to the vector of nodal temperature T(t) in the solid domain. A comparison between the newly developed code, which is finite volume based and a standard finite element-based code such as ANSYS is presented. Despite the significant reduction in CPU time, the finite-volume code well approximates the solid temperature field computed by ANSYS for the two cases considered here, i.e. the fusion power operation and vacuum vessel baking operation.
{"title":"Linear model responses in forced flow cooling","authors":"Giacomo Cavuoti , Francesca Cau , José Lorenzo , Alfredo Portone","doi":"10.1016/j.fusengdes.2025.115610","DOIUrl":"10.1016/j.fusengdes.2025.115610","url":null,"abstract":"<div><div>The aim of this paper is to present a fast method capable of computing thermo-hydraulic transients in solid components that are cooled (or heated) by incompressible forced flow with or without external heating sources. By coupling the heat conduction equation in the solid volume to the heat transfer to the forced flow we derive a linearized mapping between the vector of input (control) quantities <strong>u</strong>(t) such as mass flow and inlet temperature to the vector of nodal temperature <strong>T</strong>(t) in the solid domain. A comparison between the newly developed code, which is finite volume based and a standard finite element-based code such as ANSYS is presented. Despite the significant reduction in CPU time, the finite-volume code well approximates the solid temperature field computed by ANSYS for the two cases considered here, i.e. the fusion power operation and vacuum vessel baking operation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115610"},"PeriodicalIF":2.0,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145963132","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}
Pub Date : 2026-04-01Epub Date: 2026-02-12DOI: 10.1016/j.fusengdes.2026.115666
MinSoo Cha , MinHo Woo , Sang-hee Hahn , Mi Joung , Yong-Su Na
We report the feasibility of minimum island width growth rate seeking controller, recently implemented in Korea Superconducting Tokamak Advanced Research (KSTAR) plasma control system (PCS), and verified for the active feedback control of tearing mode (TM). Prior to experimental application, the algorithm needs to be validated under actual experimental conditions. To this end, we extended the TokSys simulator to model both electron cyclotron (EC) beam steering dynamics and TM evolution. The TokSys simulator was originally developed to enhance axisymmetric magnetic control and simulate the dynamic responses of plasma. The newly implemented modules include the EC response simulator, EC current drive (ECCD) modelling module, modified Rutherford equation (MRE) solver, and Mirnov diagnostics module. By conducting TokSys simulations in conjunction with the KSTAR PCS, we demonstrate that the minimum seeking algorithm can successfully stabilize the TM even under non-ideal experimental conditions. To explore this further, we analyzed the effects of tuning parameters on the TM stabilizing efficiency. Our findings suggest that real-time diagnostics of the rational surface location are more effective than estimating it from TM response alone, as such diagnostics can mitigate efficiency degradation caused by experimental adoptions such as boxcar averaging and scan phase. Nevertheless, our simulations indicate that the TM can be fully stabilized within 1 s using this algorithm and with the current EC mirror hardware and available power, paving the way for high-performance operation without TMs in KSTAR.
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