Pub Date : 2025-12-08DOI: 10.1016/j.fusengdes.2025.115578
P. Molina-Cabrera, F. Pastore, A. Frank, L. Simons, A. Tourneur, C. Yildiz, B. Vincent, K. Verhaegh , C. Marini , M. Wensing, A. Ianchenko, A. Balestri, S. Ernst, S. Coda, U. Sheikh, TCV team
As modern fusion experiments continue to push the boundaries of fusion science, the number, complexity, and importance of standard diagnostics have increased. Ensuring the recording of high-quality data from standard diagnostics is a task of great importance, entrusted to Ph.D. students in the TCV tokamak. Students participate in the control room team as the ‘diagnostician of the day’ or diagnosticien du jour (DdJ). This paper presents recent improvements to the DdJ software routines that prepare standard diagnostic settings, display, and automatically monitor the quality of diagnostic data. Recent updates have automated gain preparation in several standard diagnostics, which has led to reduced saturation and minimized signal-to-noise losses in the digitization process. Refactoring has also brought important runtime improvements to automatic data check routines. Lastly, new gain-preparation routines have been implemented that predict changes in plasma temperature due to changes in external electron heating power to better prepare the Thomson Scattering diagnostic, resulting in reduced saturation compared with traditional gain-preparation routines. These improvements have been led by a multi-generational task force: the DdJ-Ninjas.
{"title":"Improvements to standard diagnostic preparation and data-quality monitoring in the TCV tokamak","authors":"P. Molina-Cabrera, F. Pastore, A. Frank, L. Simons, A. Tourneur, C. Yildiz, B. Vincent, K. Verhaegh , C. Marini , M. Wensing, A. Ianchenko, A. Balestri, S. Ernst, S. Coda, U. Sheikh, TCV team","doi":"10.1016/j.fusengdes.2025.115578","DOIUrl":"10.1016/j.fusengdes.2025.115578","url":null,"abstract":"<div><div>As modern fusion experiments continue to push the boundaries of fusion science, the number, complexity, and importance of standard diagnostics have increased. Ensuring the recording of high-quality data from standard diagnostics is a task of great importance, entrusted to Ph.D. students in the TCV tokamak. Students participate in the control room team as the ‘diagnostician of the day’ or <em>diagnosticien du jour</em> (DdJ). This paper presents recent improvements to the DdJ software routines that prepare standard diagnostic settings, display, and automatically monitor the quality of diagnostic data. Recent updates have automated gain preparation in several standard diagnostics, which has led to reduced saturation and minimized signal-to-noise losses in the digitization process. Refactoring has also brought important runtime improvements to automatic data check routines. Lastly, new gain-preparation routines have been implemented that predict changes in plasma temperature due to changes in external electron heating power to better prepare the Thomson Scattering diagnostic, resulting in reduced saturation compared with traditional gain-preparation routines. These improvements have been led by a multi-generational task force: the DdJ-Ninjas.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115578"},"PeriodicalIF":2.0,"publicationDate":"2025-12-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145749739","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 : 2025-12-08DOI: 10.1016/j.fusengdes.2025.115588
Enrico Emanuelli , Francesco Vannini , Matthias Hoelzl , Nina Schwarz , Eric Nardon , Vinodh Bandaru , Daniele Bonfiglio , Artur Kryzhanovskyy , Giuseppe Ramogida , Fabio Subba , JOREK Team
Formation of Runaway electrons (REs) during tokamak disruptions is a significant challenge in fusion research, as they can locally damage the plasma-facing components by applying thermal loads of tens of MJ per square meter, possibly leading to significant melting. This work investigates the current quench phase of disruptions and the likelihood of RE generation and multiplication in the Day-0 scenario (plasma current MA) of the Divertor Tokamak Test (DTT), using the non-linear magnetohydrodynamic code JOREK. Our results from 2D (toroidally symmetric) simulations indicate that, in this initial low-current scenario, RE generation is minimal to negligible when the impurities injected through disruption mitigation systems are adequately limited. This suggests that DTT’s early operational phase poses a low RE risk, contributing to operational safety in this regard before transitioning to full power scenarios ( MA). In addition to providing an initial RE safety benchmark for DTT, this study lays the groundwork for further research at higher operational currents and for the estimation of heat loads caused by RE beams on plasma-facing components, essential for guiding the design and strategic placement of mitigation elements such as sacrificial limiters.
{"title":"Simulation of runaway electron generation in the Day-0 scenario of DTT","authors":"Enrico Emanuelli , Francesco Vannini , Matthias Hoelzl , Nina Schwarz , Eric Nardon , Vinodh Bandaru , Daniele Bonfiglio , Artur Kryzhanovskyy , Giuseppe Ramogida , Fabio Subba , JOREK Team","doi":"10.1016/j.fusengdes.2025.115588","DOIUrl":"10.1016/j.fusengdes.2025.115588","url":null,"abstract":"<div><div>Formation of Runaway electrons (REs) during tokamak disruptions is a significant challenge in fusion research, as they can locally damage the plasma-facing components by applying thermal loads of tens of MJ per square meter, possibly leading to significant melting. This work investigates the current quench phase of disruptions and the likelihood of RE generation and multiplication in the Day-0 scenario (plasma current <span><math><mrow><msub><mrow><mi>I</mi></mrow><mrow><mi>p</mi></mrow></msub><mo>=</mo><mn>2</mn></mrow></math></span> MA) of the Divertor Tokamak Test (DTT), using the non-linear magnetohydrodynamic code JOREK. Our results from 2D (toroidally symmetric) simulations indicate that, in this initial low-current scenario, RE generation is minimal to negligible when the impurities injected through disruption mitigation systems are adequately limited. This suggests that DTT’s early operational phase poses a low RE risk, contributing to operational safety in this regard before transitioning to full power scenarios (<span><math><mrow><msub><mrow><mi>I</mi></mrow><mrow><mi>p</mi></mrow></msub><mo>=</mo><mn>5</mn><mo>.</mo><mn>5</mn></mrow></math></span> MA). In addition to providing an initial RE safety benchmark for DTT, this study lays the groundwork for further research at higher operational currents and for the estimation of heat loads caused by RE beams on plasma-facing components, essential for guiding the design and strategic placement of mitigation elements such as sacrificial limiters.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115588"},"PeriodicalIF":2.0,"publicationDate":"2025-12-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145749741","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 : 2025-12-07DOI: 10.1016/j.fusengdes.2025.115579
Seung-Ju Lee , Myungkyu Kim , Jaesic Hong , Sang-won Yun , Taegu Lee , Sang-hee Hahn , Woong-Ryol Lee , Taehyun Tak
The KSTAR Fast Interlock System (FIS) has the primary role of protecting the devices installed in the vacuum vessel of tokamak such as Plasma Facing Components (PFCs) by immediately stopping the KSTAR heating devices, following the event handling actions of the Plasma Control System (PCS). Furthermore, the FIS assists the PCS event handling operations by redundantly detecting abnormal Plasma Current (IP) events. The initially implemented detection logic for the IP minimum fault event has been successfully evaluated and operated. In this paper, we implement another logic detecting the IP error fault event that the discrepancy between the target IP and the measured IP exceeds the criteria. As the architecture design, we assign more complicated tasks such as the waveform generation to the host server and the error fault-checking task requiring real-time operation to the target controller. Second, the Direct Memory Access (DMA) method for data communication is adopted; thus, the target controller can conduct the detection logic and the data communication in parallel without real-time performance degradation. Third, we design proper timing of the data communication for stable operation. On the host side, we employ ITER Real-Time Framework (RTF) technology for initiating the data communication with precise timing and controlling the precise execution cycle. Finally, we apply the bypass logic to prevent conflict with the same detecting operation of the PCS. We evaluate the functionality of the IP error fault detection logic in the KSTAR plasma experiments.
{"title":"Demonstration of enhanced abnormal Plasma Current detection in KSTAR Fast Interlock System","authors":"Seung-Ju Lee , Myungkyu Kim , Jaesic Hong , Sang-won Yun , Taegu Lee , Sang-hee Hahn , Woong-Ryol Lee , Taehyun Tak","doi":"10.1016/j.fusengdes.2025.115579","DOIUrl":"10.1016/j.fusengdes.2025.115579","url":null,"abstract":"<div><div>The KSTAR Fast Interlock System (FIS) has the primary role of protecting the devices installed in the vacuum vessel of tokamak such as Plasma Facing Components (PFCs) by immediately stopping the KSTAR heating devices, following the event handling actions of the Plasma Control System (PCS). Furthermore, the FIS assists the PCS event handling operations by redundantly detecting abnormal Plasma Current (IP) events. The initially implemented detection logic for the IP minimum fault event has been successfully evaluated and operated. In this paper, we implement another logic detecting the IP error fault event that the discrepancy between the target IP and the measured IP exceeds the criteria. As the architecture design, we assign more complicated tasks such as the waveform generation to the host server and the error fault-checking task requiring real-time operation to the target controller. Second, the Direct Memory Access (DMA) method for data communication is adopted; thus, the target controller can conduct the detection logic and the data communication in parallel without real-time performance degradation. Third, we design proper timing of the data communication for stable operation. On the host side, we employ ITER Real-Time Framework (RTF) technology for initiating the data communication with precise timing and controlling the precise execution cycle. Finally, we apply the bypass logic to prevent conflict with the same detecting operation of the PCS. We evaluate the functionality of the IP error fault detection logic in the KSTAR plasma experiments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115579"},"PeriodicalIF":2.0,"publicationDate":"2025-12-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145749737","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 : 2025-12-05DOI: 10.1016/j.fusengdes.2025.115574
Luigi Candido , Paul Barron , Colin Baus , Italo Godoy-Morison , John McGrady , Minoru Jimma , Satoshi Ogawa , Richard Pearson , Ben Raeves , Taishi Sugiyama , Jun Takamine , Jack Taylor , Luke Taylor-King , Satoshi Ueguchi , Andrew Wilson , Satoshi Konishi
The future deployment of commercial fusion energy depends on several critical factors, among which the development of a feasible, safe, and integrated breeding blanket (BB) plays a prominent role. Since the company was founded in 2019, Kyoto Fusioneering (KF) has been developing its capability in advanced blanket design and technology development, focusing efforts on the advancement of its own innovative concept known as SCYLLA (Self-Cooled Yuryo Lithium-Lead Advanced), a self-cooled lithium-lead type blanket using silicon carbide composite (SiC/SiC) as a structural material. Efforts to develop the SCYLLA design have employed a holistic approach focused on component modelling, identification of system interfaces between components and systems, and safety evaluation. In this paper, progress towards an application of the SCYLLA breeding blanket configuration, using a spherical Tokamak reactor as a reference, is reported. The description of the current architecture is provided, focusing on the main modifications to evolve the design from a pre-conceptual configuration to a more robust layout. From the point of view of interfaces and experimental R&D, a lithium-lead loop has also been developed by KF as part of its UNITY-1 facility, based in Kumiyama (Kyoto, Japan). This system includes comprehensive design and modelling of the tritium extraction unit. The chosen modelling strategy and the obtained results are reported in the paper and critically discussed.
{"title":"Preliminary design of the self-cooled lithium-lead SCYLLA blanket for a spherical tokamak","authors":"Luigi Candido , Paul Barron , Colin Baus , Italo Godoy-Morison , John McGrady , Minoru Jimma , Satoshi Ogawa , Richard Pearson , Ben Raeves , Taishi Sugiyama , Jun Takamine , Jack Taylor , Luke Taylor-King , Satoshi Ueguchi , Andrew Wilson , Satoshi Konishi","doi":"10.1016/j.fusengdes.2025.115574","DOIUrl":"10.1016/j.fusengdes.2025.115574","url":null,"abstract":"<div><div>The future deployment of commercial fusion energy depends on several critical factors, among which the development of a feasible, safe, and integrated breeding blanket (BB) plays a prominent role. Since the company was founded in 2019, Kyoto Fusioneering (KF) has been developing its capability in advanced blanket design and technology development, focusing efforts on the advancement of its own innovative concept known as SCYLLA (Self-Cooled Yuryo Lithium-Lead Advanced), a self-cooled lithium-lead type blanket using silicon carbide composite (SiC<span><math><msub><mrow></mrow><mrow><mi>f</mi></mrow></msub></math></span>/SiC) as a structural material. Efforts to develop the SCYLLA design have employed a holistic approach focused on component modelling, identification of system interfaces between components and systems, and safety evaluation. In this paper, progress towards an application of the SCYLLA breeding blanket configuration, using a spherical Tokamak reactor as a reference, is reported. The description of the current architecture is provided, focusing on the main modifications to evolve the design from a pre-conceptual configuration to a more robust layout. From the point of view of interfaces and experimental R&D, a lithium-lead loop has also been developed by KF as part of its UNITY-1 facility, based in Kumiyama (Kyoto, Japan). This system includes comprehensive design and modelling of the tritium extraction unit. The chosen modelling strategy and the obtained results are reported in the paper and critically discussed.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115574"},"PeriodicalIF":2.0,"publicationDate":"2025-12-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694588","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 : 2025-12-05DOI: 10.1016/j.fusengdes.2025.115467
Lingmin La , Lin Qin , Guanjie Liang , Lingling Wang
WTaVCr alloy coatings with different elemental contents and W substrates were selected for He ion irradiation experiments (irradiation energy 50 eV, irradiation dose of 1 × 1025 m-2, irradiation temperature 1273 K). After irradiation, the tungsten substrate showed a "fuzz" structure on the surface, while pinholes and convoluted structures appeared on the surface of WTaVCr. The irradiated WTaVCr alloy coatings exhibited a hardening phenomenon, The W25Ta23.5V20.8Cr30.6 alloy coating exhibits the lowest hardening rate, and the TEM observations of W25Ta23.5V20.8Cr30.6 show that the number density of He bubbles in the alloy was significantly lower than that of pure tungsten, which exhibits excellent resistance to irradiation.
{"title":"Irradiation resistance properties of WTaVCr alloy coatings","authors":"Lingmin La , Lin Qin , Guanjie Liang , Lingling Wang","doi":"10.1016/j.fusengdes.2025.115467","DOIUrl":"10.1016/j.fusengdes.2025.115467","url":null,"abstract":"<div><div>WTaVCr alloy coatings with different elemental contents and W substrates were selected for He ion irradiation experiments (irradiation energy 50 eV, irradiation dose of 1 × 10<sup>25</sup> m<sup>-2</sup>, irradiation temperature 1273 K). After irradiation, the tungsten substrate showed a \"fuzz\" structure on the surface, while pinholes and convoluted structures appeared on the surface of WTaVCr. The irradiated WTaVCr alloy coatings exhibited a hardening phenomenon, The W<sub>25</sub>Ta<sub>23.5</sub>V<sub>20.8</sub>Cr<sub>30.6</sub> alloy coating exhibits the lowest hardening rate, and the TEM observations of W<sub>25</sub>Ta<sub>23.5</sub>V<sub>20.8</sub>Cr<sub>30.6</sub> show that the number density of He bubbles in the alloy was significantly lower than that of pure tungsten, which exhibits excellent resistance to irradiation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115467"},"PeriodicalIF":2.0,"publicationDate":"2025-12-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145938653","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}
Permeator Against Vacuum was confirmed in 2023 as the reference technology for the tritium extraction and removal (TER) system of the Water-Cooled Lithium-Lead Breeding Blanket (WCLL BB) due to its overall better performances and higher Technology Readiness Level. The manufacturing and the first characterization of a PAV with a niobium membrane in a shell and tube configuration with U-tubes (PAV-ONE mock-up) was recently performed at ENEA Brasimone R.C., demonstrating that this technology can be satisfactorily employed in PbLi. This paper will present the design of a new PAV test section with niobium membrane to be installed in the TRIEX-II facility. The objective of the new mock-up is to investigate the correlation between the extraction flux and different parameters to optimize the future design of the technology. In particular, PAV-two will allow to deeply examine the influence of turbulence, vacuum pressure, and surface conditions on the hydrogen transport in the system and, therefore, on the performances of the technology. The simple and flexible design of PAV-two will enable the discrimination of each parameter’s impact on the extracted flux in a repeatable and reliable manner.
{"title":"PAV-2: a new mock-up to investigate niobium membrane-PAV performances optimization in PbLi systems","authors":"Francesca Papa , Ciro Alberghi , Vincenzo Claps , Daniele Martelli , Alessandro Venturini","doi":"10.1016/j.fusengdes.2025.115554","DOIUrl":"10.1016/j.fusengdes.2025.115554","url":null,"abstract":"<div><div>Permeator Against Vacuum was confirmed in 2023 as the reference technology for the tritium extraction and removal (TER) system of the Water-Cooled Lithium-Lead Breeding Blanket (WCLL BB) due to its overall better performances and higher Technology Readiness Level. The manufacturing and the first characterization of a PAV with a niobium membrane in a shell and tube configuration with U-tubes (PAV-ONE mock-up) was recently performed at ENEA Brasimone R.C., demonstrating that this technology can be satisfactorily employed in PbLi. This paper will present the design of a new PAV test section with niobium membrane to be installed in the TRIEX-II facility. The objective of the new mock-up is to investigate the correlation between the extraction flux and different parameters to optimize the future design of the technology. In particular, PAV-two will allow to deeply examine the influence of turbulence, vacuum pressure, and surface conditions on the hydrogen transport in the system and, therefore, on the performances of the technology. The simple and flexible design of PAV-two will enable the discrimination of each parameter’s impact on the extracted flux in a repeatable and reliable manner.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115554"},"PeriodicalIF":2.0,"publicationDate":"2025-12-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694590","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}
To assess the accuracy of iron data in the latest nuclear data libraries, mainly FENDL-3.2b, for accelerator-based fusion neutron source designs, we analyzed the QST/TIARA iron experiment with quasi mono-energy neutrons of 40 and 65 MeV, and the JAEA/FNS iron experiment with DT neutrons, by using the Monte Carlo code MCNP6.2. As the results, we found the following issues: 1) the calculation result with FENDL-3.2b underestimated the measured neutron flux in the continuous energy range (10 - 60 MeV) by 40 % in the TIARA experiment with 65 MeV neutrons, 2) it tended to underestimate the measured neutron flux above 10 MeV by 20 % at a depth of 70 cm and overestimate that below 10 keV by 30 % up to a depth of 40 cm in the FNS experiment. We modified the FENDL-3.2b iron data to investigate these issues and identified underlying remarks: 1) the non-elastic and inelastic scattering data of 56Fe in FENDL-3.2b underestimated the measured neutron flux above 10 MeV, 2) the (n,np) data of 56Fe in FENDL-3.2b overestimated the measured neutron flux above 10 MeV, and 3) the inelastic scattering and (n,2n) data of 56Fe and the inelastic scattering data of 57Fe in FENDL-3.2b caused the overestimation of the measured neutron flux below 10 keV. These issues of 56,57Fe in FENDL-3.2b should be improved.
{"title":"Benchmarks of iron nuclear data for fusion neutron sources","authors":"Saerom Kwon , Chikara Konno , Shogo Honda , Shunsuke Kenjo , Satoshi Sato","doi":"10.1016/j.fusengdes.2025.115548","DOIUrl":"10.1016/j.fusengdes.2025.115548","url":null,"abstract":"<div><div>To assess the accuracy of iron data in the latest nuclear data libraries, mainly FENDL-3.2b, for accelerator-based fusion neutron source designs, we analyzed the QST/TIARA iron experiment with quasi mono-energy neutrons of 40 and 65 MeV, and the JAEA/FNS iron experiment with DT neutrons, by using the Monte Carlo code MCNP6.2. As the results, we found the following issues: 1) the calculation result with FENDL-3.2b underestimated the measured neutron flux in the continuous energy range (10 - 60 MeV) by 40 % in the TIARA experiment with 65 MeV neutrons, 2) it tended to underestimate the measured neutron flux above 10 MeV by 20 % at a depth of 70 cm and overestimate that below 10 keV by 30 % up to a depth of 40 cm in the FNS experiment. We modified the FENDL-3.2b iron data to investigate these issues and identified underlying remarks: 1) the non-elastic and inelastic scattering data of <sup>56</sup>Fe in FENDL-3.2b underestimated the measured neutron flux above 10 MeV, 2) the (n,np) data of <sup>56</sup>Fe in FENDL-3.2b overestimated the measured neutron flux above 10 MeV, and 3) the inelastic scattering and (n,2n) data of <sup>56</sup>Fe and the inelastic scattering data of <sup>57</sup>Fe in FENDL-3.2b caused the overestimation of the measured neutron flux below 10 keV. These issues of <sup>56,57</sup>Fe in FENDL-3.2b should be improved.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115548"},"PeriodicalIF":2.0,"publicationDate":"2025-12-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694589","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 : 2025-12-03DOI: 10.1016/j.fusengdes.2025.115552
Sofia Bertolami , Franco Di Paolo , Alessandro Bruschi , Francesco Fanale , Alessandro Moro , Saul Garavaglia , Gustavo Granucci , Afra Romano , Alessandro Simonetto
In an Electron Cyclotron Resonance Heating (ECRH) system, to efficiently couple the signal power to the plasma, the signal wave polarization must be accurately matched to the plasma conditions at the plasma boundary. However, the millimeter-wave radiation from the power source (gyrotron) is normally linearly polarized: consequently, some kind of polarization matching is required. This study focuses on the design of a grating polarizer with sinusoidal grooves for the 170 GHz ECRH system, with an application specifically intended for the Divertor Tokamak Test (DTT), currently under construction in Frascati, Italy. To enable the generation of all possible output polarization states, a pair of polarizer mirrors will be employed and integrated into the Quasi-Optical (QO) transmission line connecting the gyrotrons to the Electron Cyclotron (EC) waves launchers. The primary objective of this study is to describe an analytical tool capable of providing detailed insights into the polarization characteristics of the reflected electric field resulting from the interaction between the incident wave and the polarizer. Additionally, the proposed program tool calculates the precise combinations of rotation angles required for the polarizers to achieve the desired output polarization states. The accuracy and reliability of the model’s prediction have been validated by comparing them with simulations conducted using commercial electromagnetic software.
{"title":"Mathematical modeling and design of a microwave polarizer for DTT ECRH applications","authors":"Sofia Bertolami , Franco Di Paolo , Alessandro Bruschi , Francesco Fanale , Alessandro Moro , Saul Garavaglia , Gustavo Granucci , Afra Romano , Alessandro Simonetto","doi":"10.1016/j.fusengdes.2025.115552","DOIUrl":"10.1016/j.fusengdes.2025.115552","url":null,"abstract":"<div><div>In an Electron Cyclotron Resonance Heating (ECRH) system, to efficiently couple the signal power to the plasma, the signal wave polarization must be accurately matched to the plasma conditions at the plasma boundary. However, the millimeter-wave radiation from the power source (gyrotron) is normally linearly polarized: consequently, some kind of polarization matching is required. This study focuses on the design of a grating polarizer with sinusoidal grooves for the 170 GHz ECRH system, with an application specifically intended for the Divertor Tokamak Test (DTT), currently under construction in Frascati, Italy. To enable the generation of all possible output polarization states, a pair of polarizer mirrors will be employed and integrated into the Quasi-Optical (QO) transmission line connecting the gyrotrons to the Electron Cyclotron (EC) waves launchers. The primary objective of this study is to describe an analytical tool capable of providing detailed insights into the polarization characteristics of the reflected electric field resulting from the interaction between the incident wave and the polarizer. Additionally, the proposed program tool calculates the precise combinations of rotation angles required for the polarizers to achieve the desired output polarization states. The accuracy and reliability of the model’s prediction have been validated by comparing them with simulations conducted using commercial electromagnetic software.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115552"},"PeriodicalIF":2.0,"publicationDate":"2025-12-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145685390","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 : 2025-12-03DOI: 10.1016/j.fusengdes.2025.115550
A.A. Shoshin
The description of the main regulatory documents applied in the design and construction of the elements of the international thermonuclear reactor ITER in France is given, their main requirements are presented. Significant difficulties with the design and manufacture of components arise because ITER is a nuclear facility under French law. The French classification of pressure equipment (otherwise called 'pressurized equipment') in nuclear facilities is considered, examples of ITER diagnostic port equipment are given. The difficulties arising from the application of these regulatory documents are shown. The main rules and requirements developed by the ITER Organization itself for vacuum equipment and mechanical components are listed. The main industry standards used in this project are reviewed. One possible solution that could facilitate the development and construction of fusion reactors is to develop regulations specifically for fusion plants.
{"title":"Directives, codes, standards and other requirements applicable to the design and manufacture of components in the ITER project","authors":"A.A. Shoshin","doi":"10.1016/j.fusengdes.2025.115550","DOIUrl":"10.1016/j.fusengdes.2025.115550","url":null,"abstract":"<div><div>The description of the main regulatory documents applied in the design and construction of the elements of the international thermonuclear reactor ITER in France is given, their main requirements are presented. Significant difficulties with the design and manufacture of components arise because ITER is a nuclear facility under French law. The French classification of pressure equipment (otherwise called 'pressurized equipment') in nuclear facilities is considered, examples of ITER diagnostic port equipment are given. The difficulties arising from the application of these regulatory documents are shown. The main rules and requirements developed by the ITER Organization itself for vacuum equipment and mechanical components are listed. The main industry standards used in this project are reviewed. One possible solution that could facilitate the development and construction of fusion reactors is to develop regulations specifically for fusion plants.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115550"},"PeriodicalIF":2.0,"publicationDate":"2025-12-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694586","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 : 2025-12-02DOI: 10.1016/j.fusengdes.2025.115547
Young Ah Park , Yi-Hyun Park , Mu-Young Ahn , Young Soo Yoon
Li4SiO4, used as a tritium breeder material in fusion reactors, has high lithium density but suffers from lower mechanical strength compared to another promising material, Li2TiO3. In this study, a fluidic system was designed to fabricate core-shell structured pebbles (Li4SiO4–Li2TiO3) consisting of a Li4SiO4 core and a Li2TiO3 shell. The system was constructed using a T–shaped fluid flow channel with a double–tube design, enabling the controlled formation of core-shell droplets by adjusting the flow rates of the continuous and dispersed phases. The resulting Li4SiO4–Li2TiO3 pebbles exhibited a crush load approximately 2.03 times higher than that of single-phase Li4SiO4 pebbles, and the stable formation of the core–shell structure was confirmed. This study presents a novel fabrication process with the potential to enhance the mechanical performance of Li4SiO4 as a tritium breeder material.
{"title":"Preliminary study on the development of the fluidic system for the fabrication of pebble with core–shell structure","authors":"Young Ah Park , Yi-Hyun Park , Mu-Young Ahn , Young Soo Yoon","doi":"10.1016/j.fusengdes.2025.115547","DOIUrl":"10.1016/j.fusengdes.2025.115547","url":null,"abstract":"<div><div>Li<sub>4</sub>SiO<sub>4</sub>, used as a tritium breeder material in fusion reactors, has high lithium density but suffers from lower mechanical strength compared to another promising material, Li<sub>2</sub>TiO<sub>3</sub>. In this study, a fluidic system was designed to fabricate core-shell structured pebbles (Li<sub>4</sub>SiO<sub>4</sub>–Li<sub>2</sub>TiO<sub>3</sub>) consisting of a Li<sub>4</sub>SiO<sub>4</sub> core and a Li<sub>2</sub>TiO<sub>3</sub> shell. The system was constructed using a T–shaped fluid flow channel with a double–tube design, enabling the controlled formation of core-shell droplets by adjusting the flow rates of the continuous and dispersed phases. The resulting Li<sub>4</sub>SiO<sub>4</sub>–Li<sub>2</sub>TiO<sub>3</sub> pebbles exhibited a crush load approximately 2.03 times higher than that of single-phase Li<sub>4</sub>SiO<sub>4</sub> pebbles, and the stable formation of the core–shell structure was confirmed. This study presents a novel fabrication process with the potential to enhance the mechanical performance of Li<sub>4</sub>SiO<sub>4</sub> as a tritium breeder material.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115547"},"PeriodicalIF":2.0,"publicationDate":"2025-12-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694585","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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