Pub 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":"2026-01-21","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 : 2026-01-21DOI: 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":"2026-01-21","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 : 2026-01-20DOI: 10.1016/j.fusengdes.2026.115626
M.P. Ross , K.A. Thackston , A. Dupuy , Y. Gorelov , N. de Boucaud , P. Nesbet , A. Torrezan , Z. Bayler , N. Watson , J. Anderson , J.P. Squire
Electron cyclotron heating (ECH) and current drive (ECCD) will play a large role in tokamak-based fusion reactors. At the DIII-D tokamak, 110 GHz microwaves injected into the plasma can provide core heating and current drive as well as impurity control, neoclassical tearing mode mitigation, and breakdown assistance. Understanding the physics of these processes relies on accurate estimates of injected ECH power. DIII-D’s ECH system consists of six MW-class Microwave Power Products (MPP) gyrotron microwave sources. Operating the gyrotrons far from the tokamak removes them from magnetic field interference, so 31.75 mm inner-diameter corrugated waveguides transmit the microwave power the 80 m from the gyrotrons to steerable launchers in the tokamak chamber. Estimates of injected power rely on knowing the generated power at the source and then subtracting transmission loss. Conventional transmission loss measurements based on calorimetric dummy loads are onerous and only possible during extended maintenance periods. This work examines two tools that provide more flexibility for the transmission loss measurements. A resistive temperature detector (RTD) array installed along a waveguide measures heat lost to the transmission line, and low power time domain reflectometry (TDR) measurements with a vector network analyzer (VNA) allows loss measurements without burdensome hardware modifications.
{"title":"Corroborating VNA and thermal measurements of transmission loss on the DIII-D ECH waveguide system","authors":"M.P. Ross , K.A. Thackston , A. Dupuy , Y. Gorelov , N. de Boucaud , P. Nesbet , A. Torrezan , Z. Bayler , N. Watson , J. Anderson , J.P. Squire","doi":"10.1016/j.fusengdes.2026.115626","DOIUrl":"10.1016/j.fusengdes.2026.115626","url":null,"abstract":"<div><div>Electron cyclotron heating (ECH) and current drive (ECCD) will play a large role in tokamak-based fusion reactors. At the DIII-D tokamak, 110 GHz microwaves injected into the plasma can provide core heating and current drive as well as impurity control, neoclassical tearing mode mitigation, and breakdown assistance. Understanding the physics of these processes relies on accurate estimates of injected ECH power. DIII-D’s ECH system consists of six MW-class Microwave Power Products (MPP) gyrotron microwave sources. Operating the gyrotrons far from the tokamak removes them from magnetic field interference, so 31.75 mm inner-diameter corrugated waveguides transmit the microwave power the 80 m from the gyrotrons to steerable launchers in the tokamak chamber. Estimates of injected power rely on knowing the generated power at the source and then subtracting transmission loss. Conventional transmission loss measurements based on calorimetric dummy loads are onerous and only possible during extended maintenance periods. This work examines two tools that provide more flexibility for the transmission loss measurements. A resistive temperature detector (RTD) array installed along a waveguide measures heat lost to the transmission line, and low power time domain reflectometry (TDR) measurements with a vector network analyzer (VNA) allows loss measurements without burdensome hardware modifications.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115626"},"PeriodicalIF":2.0,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146025282","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 : 2026-01-20DOI: 10.1016/j.fusengdes.2026.115638
K. Iwasaki , S. Sugiyama , Y. Ohtani , Y. Sakamoto
The toroidal interferometer and polarimeter (TIP) have been investigated as density diagnostics for JA DEMO. A model for the interferometer and polarimeter phase shifts incorporating the finite electron temperature effect has been implemented into a plasma control simulation code to generate the synthetic phase shift signals. The laser wavelength is set to m identical to that used in ITER. The finite temperature effect is significant. The deviation between the estimated line-averaged densities obtained with and without accounting for finite temperature effects reaches approximately 7% for the interferometer and 10% for the polarimeter along lines of sight near the magnetic axis, and decreases to a few percent near the outer edge. Density feedback control has been performed, and a comparison is made between the line-averaged densities with and without correction for the temperature effect. When the temperature effect is neglected, the density is underestimated, leading to an increase in the actual density. Consequently, the fusion output increases, resulting in an error of up to 11% when using the central viewing chords. Correction of the density error caused by the finite electron temperature has been carried out using TIP alone by taking the difference between the interferometer and polarimeter signals. The results show that it can reduce the density error to below 1%.
{"title":"Consideration of density measurement using toroidal interferometer and polarimeter on JA DEMO","authors":"K. Iwasaki , S. Sugiyama , Y. Ohtani , Y. Sakamoto","doi":"10.1016/j.fusengdes.2026.115638","DOIUrl":"10.1016/j.fusengdes.2026.115638","url":null,"abstract":"<div><div>The toroidal interferometer and polarimeter (TIP) have been investigated as density diagnostics for JA DEMO. A model for the interferometer and polarimeter phase shifts incorporating the finite electron temperature effect has been implemented into a plasma control simulation code to generate the synthetic phase shift signals. The laser wavelength is set to <span><math><mrow><mn>10</mn><mo>.</mo><mn>6</mn><mspace></mspace><mi>μ</mi></mrow></math></span>m identical to that used in ITER. The finite temperature effect is significant. The deviation between the estimated line-averaged densities obtained with and without accounting for finite temperature effects reaches approximately 7% for the interferometer and 10% for the polarimeter along lines of sight near the magnetic axis, and decreases to a few percent near the outer edge. Density feedback control has been performed, and a comparison is made between the line-averaged densities with and without correction for the temperature effect. When the temperature effect is neglected, the density is underestimated, leading to an increase in the actual density. Consequently, the fusion output increases, resulting in an error of up to 11% when using the central viewing chords. Correction of the density error caused by the finite electron temperature has been carried out using TIP alone by taking the difference between the interferometer and polarimeter signals. The results show that it can reduce the density error to below 1%.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115638"},"PeriodicalIF":2.0,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146025233","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 : 2026-01-20DOI: 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":"2026-01-20","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 : 2026-01-19DOI: 10.1016/j.fusengdes.2026.115634
Jingquan Xia , Zhilin Chen , Yang Yang , Zhu An
The tritium depth distribution in the structural material (i.e., the 5083 aluminum alloy (Al5083)) of the inertial confinement fusion (ICF) target chamber is a key basis for evaluating radiation protection and environmental hazards. Given the characteristic of shallow tritium ion implantation depth in the Al5083, this study aimed to establish a high-resolution measurement method. A method for measuring the tritium depth distribution in Al5083 was established by coupling the ZnCl2 surface washing (SW) method and the layer-by-layer chemical etching (LLCE) method. Firstly, the effects of etchant type, acid-to-water ratio, temperature, and etching period on measurement resolution in the LLCE method were systematically investigated, and the depth resolution was controlled within 30–50 nm, meeting the requirement for high-resolution measurement of shallowly implanted tritium. Secondly, the reliability of two correction methods for liquid scintillation counting (LSC) measurement of tritium-containing zinc solution was verified using the internal standard method. It was confirmed that the SW method can accurately separate and quantitatively determine the total amount of surface-adsorbed tritium in the Al5083, providing a key method to distinguish between surface-adsorbed tritium and bulk-retained tritium. Experiments on tritium-loaded Al5083 sample by implantation with an average incident energy of 7.5 keV and a fluence of 4.5 × 10¹² ions/cm² showed that the surface-adsorbed tritium accounts for 1.8% of the total tritium amount with a high concentration (∼108 Bq/cm3), and 80% of the tritium in the bulk is distributed within 200 nm of the surface. The standardized measurement method established in this study provides reliable technical support for evaluating the retention and migration of tritium in the structural material of the ICF device.
{"title":"Tritium depth distribution measurement by using chemical etching and surface washing technique for structural material of the ICF target chamber","authors":"Jingquan Xia , Zhilin Chen , Yang Yang , Zhu An","doi":"10.1016/j.fusengdes.2026.115634","DOIUrl":"10.1016/j.fusengdes.2026.115634","url":null,"abstract":"<div><div>The tritium depth distribution in the structural material (i.e., the 5083 aluminum alloy (Al5083)) of the inertial confinement fusion (ICF) target chamber is a key basis for evaluating radiation protection and environmental hazards. Given the characteristic of shallow tritium ion implantation depth in the Al5083, this study aimed to establish a high-resolution measurement method. A method for measuring the tritium depth distribution in Al5083 was established by coupling the ZnCl<sub>2</sub> surface washing (SW) method and the layer-by-layer chemical etching (LLCE) method. Firstly, the effects of etchant type, acid-to-water ratio, temperature, and etching period on measurement resolution in the LLCE method were systematically investigated, and the depth resolution was controlled within 30–50 nm, meeting the requirement for high-resolution measurement of shallowly implanted tritium. Secondly, the reliability of two correction methods for liquid scintillation counting (LSC) measurement of tritium-containing zinc solution was verified using the internal standard method. It was confirmed that the SW method can accurately separate and quantitatively determine the total amount of surface-adsorbed tritium in the Al5083, providing a key method to distinguish between surface-adsorbed tritium and bulk-retained tritium. Experiments on tritium-loaded Al5083 sample by implantation with an average incident energy of 7.5 keV and a fluence of 4.5 × 10¹² ions/cm² showed that the surface-adsorbed tritium accounts for 1.8% of the total tritium amount with a high concentration (∼10<sup>8</sup> Bq/cm<sup>3</sup>), and 80% of the tritium in the bulk is distributed within 200 nm of the surface. The standardized measurement method established in this study provides reliable technical support for evaluating the retention and migration of tritium in the structural material of the ICF device.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115634"},"PeriodicalIF":2.0,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146025284","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 : 2026-01-15DOI: 10.1016/j.fusengdes.2026.115630
Feng Zhang, Mei Huang, Gangyu Chen, He Wang, Wanxin Zheng, Jieqiong Wang, Guoyao Fan, Cheng Chen
As a crucial component of the electron cyclotron resonance heating (ECRH) system, the polarizer primarily serves to change the polarization characteristics of millimeter wave. In this study, an ultra-wideband polarization strategy based on two polarizers for the ECRH system is presented. By employing two identical sinusoidally grooved polarizers at a designated incident angle of 60°, it is possible to attain arbitrary polarization can be attained across an ultra-wideband frequency ranging from 99 GHz to 189 GHz. A ultra-wideband polarizer was devised and evaluated, and computational results indicate that the arbitrary polarization efficiency of the proposed method reaches at least 99.94%. According to this analysis, nearly every desired polarization state can be realized using the presented polarization strategy.
{"title":"Investigation of the ultra-wideband polarizer for high power millimeter wave system","authors":"Feng Zhang, Mei Huang, Gangyu Chen, He Wang, Wanxin Zheng, Jieqiong Wang, Guoyao Fan, Cheng Chen","doi":"10.1016/j.fusengdes.2026.115630","DOIUrl":"10.1016/j.fusengdes.2026.115630","url":null,"abstract":"<div><div>As a crucial component of the electron cyclotron resonance heating (ECRH) system, the polarizer primarily serves to change the polarization characteristics of millimeter wave. In this study, an ultra-wideband polarization strategy based on two polarizers for the ECRH system is presented. By employing two identical sinusoidally grooved polarizers at a designated incident angle of 60°, it is possible to attain arbitrary polarization can be attained across an ultra-wideband frequency ranging from 99 GHz to 189 GHz. A ultra-wideband polarizer was devised and evaluated, and computational results indicate that the arbitrary polarization efficiency of the proposed method reaches at least 99.94%. According to this analysis, nearly every desired polarization state can be realized using the presented polarization strategy.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115630"},"PeriodicalIF":2.0,"publicationDate":"2026-01-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145963133","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 : 2026-01-15DOI: 10.1016/j.fusengdes.2026.115629
Haixia Wang , Xuewei Fu , Zihui Yang , Qianchao Huo , Jie Yu
The China Fusion Engineering Test Reactor (CFETR) is a next-generation fusion reactor project independently designed and actively pursued by China. As the key radioactive source term of the CFETR, tritium safety is an important issue of nuclear safety. The tritium safety confinement system is one of the major systems in the CFETR tritium plant, and the study of tritium transport behavior in confinement system is of significant research interest. Supported by the National Key R&D Program, this study employs Unreal Engine (UE) as the 3D interactive simulation engine to construct a 3D simulation demonstration platform for tritium safety confinement of the CFETR. The Tokamak Exhaust Processing (TEP) System is selected as a representative case for simulation. Test results indicate that the platform enables smooth human-computer interaction, effectively visualizes tritium transport behavior under typical conditions, and provides an immersive 3D scene virtual roaming experience. Through dynamic demonstration of tritium transport under multiple conditions, the design principle of the CFETR tritium safety confinement system is effectively visualized, offering valuable insights for the future design of tritium confinement system in tritium plants.
{"title":"Development of 3D simulation demonstration platform for tritium safety confinement of China fusion engineering test reactor","authors":"Haixia Wang , Xuewei Fu , Zihui Yang , Qianchao Huo , Jie Yu","doi":"10.1016/j.fusengdes.2026.115629","DOIUrl":"10.1016/j.fusengdes.2026.115629","url":null,"abstract":"<div><div>The China Fusion Engineering Test Reactor (CFETR) is a next-generation fusion reactor project independently designed and actively pursued by China. As the key radioactive source term of the CFETR, tritium safety is an important issue of nuclear safety. The tritium safety confinement system is one of the major systems in the CFETR tritium plant, and the study of tritium transport behavior in confinement system is of significant research interest. Supported by the National Key R&D Program, this study employs Unreal Engine (UE) as the 3D interactive simulation engine to construct a 3D simulation demonstration platform for tritium safety confinement of the CFETR. The Tokamak Exhaust Processing (TEP) System is selected as a representative case for simulation. Test results indicate that the platform enables smooth human-computer interaction, effectively visualizes tritium transport behavior under typical conditions, and provides an immersive 3D scene virtual roaming experience. Through dynamic demonstration of tritium transport under multiple conditions, the design principle of the CFETR tritium safety confinement system is effectively visualized, offering valuable insights for the future design of tritium confinement system in tritium plants.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115629"},"PeriodicalIF":2.0,"publicationDate":"2026-01-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981155","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 : 2026-01-15DOI: 10.1016/j.fusengdes.2026.115628
Qin Lei , Qiannan Yu , Jiaguan Peng , Yiwen Sun , Mengqi Zhang , Hanfeng Song , Hao Yin , Xiaolu Xiong , Sijie Hao , Yuhao Li , Xiuli Zhu , Lu Sun , Long Cheng , Yue Yuan , Guang-Hong Lu
The study of hydrogen isotopes (HIs) retention in fusion materials is crucial for the safe operation of fusion devices. This research developed the proximal probe thermal desorption mass spectrometry (PTDS) technique, a unique technique utilizing a probe to heat the specific micro-regions on the sample, to characterize lateral deuterium (D) retention in materials using deuterated tungsten. The temperature during the test was studied using an infrared camera, with the maximum probe tip temperature reaching approximately 2300 K and the temperature-affected region having a diameter of about 500 µm. PTDS testing of deuterated tungsten films prepared by magnetron sputtering revealed a uniform lateral distribution of D retention on the sample surface. Using PTDS, the estimated deuterium-to-tungsten (D/W) atomic ratio was approximately 0.146. PTDS testing of D plasma-exposed sample provided the lateral distribution characteristics of D retention, which were compared with the distribution of plasma flux showing consistent trends. Besides, D/W ratio in D plasma-exposed sample is about 6.40 × 10–5 as estimated based on thermal desorption spectroscopy (TDS) measurement, indicating that the detection capability of the device reached the order of ∼10–5. Furthermore, this technique is of potential in precise localization of measurement spots and future research will focus on enhancing the lateral spatial resolution and in situ application of PTDS on HIs transport studies during plasma exposure.
{"title":"Proximal probe thermal desorption mass spectrometry for mapping lateral hydrogen isotope retention in metals","authors":"Qin Lei , Qiannan Yu , Jiaguan Peng , Yiwen Sun , Mengqi Zhang , Hanfeng Song , Hao Yin , Xiaolu Xiong , Sijie Hao , Yuhao Li , Xiuli Zhu , Lu Sun , Long Cheng , Yue Yuan , Guang-Hong Lu","doi":"10.1016/j.fusengdes.2026.115628","DOIUrl":"10.1016/j.fusengdes.2026.115628","url":null,"abstract":"<div><div>The study of hydrogen isotopes (HIs) retention in fusion materials is crucial for the safe operation of fusion devices. This research developed the proximal probe thermal desorption mass spectrometry (PTDS) technique, a unique technique utilizing a probe to heat the specific micro-regions on the sample, to characterize lateral deuterium (D) retention in materials using deuterated tungsten. The temperature during the test was studied using an infrared camera, with the maximum probe tip temperature reaching approximately 2300 K and the temperature-affected region having a diameter of about 500 µm. PTDS testing of deuterated tungsten films prepared by magnetron sputtering revealed a uniform lateral distribution of D retention on the sample surface. Using PTDS, the estimated deuterium-to-tungsten (D/W) atomic ratio was approximately 0.146. PTDS testing of D plasma-exposed sample provided the lateral distribution characteristics of D retention, which were compared with the distribution of plasma flux showing consistent trends. Besides, D/W ratio in D plasma-exposed sample is about 6.40 × 10<sup>–5</sup> as estimated based on thermal desorption spectroscopy (TDS) measurement, indicating that the detection capability of the device reached the order of ∼10<sup>–5</sup>. Furthermore, this technique is of potential in precise localization of measurement spots and future research will focus on enhancing the lateral spatial resolution and <em>in situ</em> application of PTDS on HIs transport studies during plasma exposure.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"225 ","pages":"Article 115628"},"PeriodicalIF":2.0,"publicationDate":"2026-01-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145963131","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 : 2026-01-14DOI: 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":"2026-01-14","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}
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