Pub Date : 2025-12-11DOI: 10.1016/j.fusengdes.2025.115577
R. Castro , Y. Makushok , L. Abadie , J. Vega
ITER, one of the most advanced fusion projects, requires handling massive amounts of data generated in real time and stored in distributed repositories. The diversity in the nature of the data, from control variables to fast acquisition signals, poses significant challenges for efficient access and organization of the information. This paper presents a large-scale indexing system designed to meet these needs. The system, integrated into ITER's CODAC core, implements a distributed and scalable architecture that ensures real-time indexing, robustness, and fault tolerance. Its design, implementation and performance are described here, highlighting its capacity to handle more than one petabyte of data per day and respond in real time to user and system requests. This breakthrough contributes significantly to efficient data handling in long-lived fusion environments.
{"title":"Large-scale indexing system for ITER data handling","authors":"R. Castro , Y. Makushok , L. Abadie , J. Vega","doi":"10.1016/j.fusengdes.2025.115577","DOIUrl":"10.1016/j.fusengdes.2025.115577","url":null,"abstract":"<div><div>ITER, one of the most advanced fusion projects, requires handling massive amounts of data generated in real time and stored in distributed repositories. The diversity in the nature of the data, from control variables to fast acquisition signals, poses significant challenges for efficient access and organization of the information. This paper presents a large-scale indexing system designed to meet these needs. The system, integrated into ITER's CODAC core, implements a distributed and scalable architecture that ensures real-time indexing, robustness, and fault tolerance. Its design, implementation and performance are described here, highlighting its capacity to handle more than one petabyte of data per day and respond in real time to user and system requests. This breakthrough contributes significantly to efficient data handling in long-lived fusion environments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"224 ","pages":"Article 115577"},"PeriodicalIF":2.0,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145718900","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}
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}
Pub Date : 2025-12-02DOI: 10.1016/j.fusengdes.2025.115527
Ch. Klein, F. Arbeiter, M Enke
The HCPB (Helium Cooled Pebble Bed) blanket concept for EU DEMO fusion reactors employs high-pressure (8 MPa) helium gas as coolant for the plasma facing first wall (FW). Up-to-date estimates for the total maximum of heat flux go up to 0.73 MW/m². Higher short-term transient loads are possible. 60°-V-shaped ribs show high heat transfer and are thus the subject of the presented studies. Although Scale-Resolving Simulation (SRS) techniques such as LES (Large Eddy Simulation) are able to calculate heat transfer and pressure drop precisely, their application is limited to an individual ribs or a few mm channel segment due to the high required mesh count. Nevertheless, SRS techniques can be used to compare different ribs and to evaluate the performance of RANS (Reynolds-Averaged Navier-Stokes Simulations). Selected RANS models can be used to evaluate the development of secondary flow structures along the channel and compare complete channels with different rib configurations for thermohydraulic performance.
The objectives of the present paper are (i) to offer guidance on the range and limits of applicability of numerical methods when dealing with cooling flows in structured channels (sections 1–4) and (ii) to provide results on specific design features of surface structures (sections 5–7) that help designers in implementing thermal-hydraulic efficient yet fabrication friendly structured channels
Challenges like long entrance length and high material properties gradients are shown. Strategies for ribs height reduction with increasing heat transfer and reduction of pressure drop resulting in higher Cooling Performance Numbers (CPN) are found. Thermohydraulic performance of fabrication friendly ribs is calculated along the channel.
用于EU DEMO聚变反应堆的HCPB（氦冷却球床）包层概念采用高压（8 MPa）氦气作为面向第一壁（FW）的等离子体的冷却剂。最新估计的最大总热通量高达0.73 MW/m²。更高的短期暂态负载是可能的。60°v形肋具有高的传热性能，因此是本研究的主题。尽管像LES（大涡模拟）这样的尺度解析模拟（SRS）技术能够精确地计算传热和压降，但由于需要很高的网格数，它们的应用仅限于单个肋或几毫米的通道段。然而，SRS技术可以用来比较不同的肋，并评估RANS （reynolds - average Navier-Stokes simulation）的性能。所选择的RANS模型可用于评估通道沿线二次流结构的发展，并比较不同肋形的完整通道的热水力性能。本论文的目的是(i)在处理结构通道中的冷却流动时，提供关于数值方法的适用范围和限制的指导（第1-4节）；（ii）提供表面结构的特定设计特征的结果（第5-7节），帮助设计师实现热水力高效且制造友好的结构通道。研究了降低肋部高度、增加传热和降低压降从而提高冷却性能数值的策略。沿通道计算了加工友好肋的热工性能。
{"title":"Numerical analysis of 60° V-ribs for helium-cooled high heat flux loaded first wall, challenges and contributions","authors":"Ch. Klein, F. Arbeiter, M Enke","doi":"10.1016/j.fusengdes.2025.115527","DOIUrl":"10.1016/j.fusengdes.2025.115527","url":null,"abstract":"<div><div>The HCPB (Helium Cooled Pebble Bed) blanket concept for EU DEMO fusion reactors employs high-pressure (8 MPa) helium gas as coolant for the plasma facing first wall (FW). Up-to-date estimates for the total maximum of heat flux go up to 0.73 MW/m². Higher short-term transient loads are possible. 60°-V-shaped ribs show high heat transfer and are thus the subject of the presented studies. Although Scale-Resolving Simulation (SRS) techniques such as LES (Large Eddy Simulation) are able to calculate heat transfer and pressure drop precisely, their application is limited to an individual ribs or a few mm channel segment due to the high required mesh count. Nevertheless, SRS techniques can be used to compare different ribs and to evaluate the performance of RANS (Reynolds-Averaged Navier-Stokes Simulations). Selected RANS models can be used to evaluate the development of secondary flow structures along the channel and compare complete channels with different rib configurations for thermohydraulic performance.</div><div>The objectives of the present paper are (i) to offer guidance on the range and limits of applicability of numerical methods when dealing with cooling flows in structured channels (sections 1–4) and (ii) to provide results on specific design features of surface structures (sections 5–7) that help designers in implementing thermal-hydraulic efficient yet fabrication friendly structured channels</div><div>Challenges like long entrance length and high material properties gradients are shown. Strategies for ribs height reduction with increasing heat transfer and reduction of pressure drop resulting in higher Cooling Performance Numbers (CPN) are found. Thermohydraulic performance of fabrication friendly ribs is calculated along the channel.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115527"},"PeriodicalIF":2.0,"publicationDate":"2025-12-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694773","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}
Significant quantities of neutrons and gamma rays deduce the nuclear heating on the shielding walls of the test cell (TC) in the A-FNS. This study presents nuclear and thermal analyses of the shielding wall in the A-FNS to evaluate its radiation-shielding effectiveness and cooling capacity. Previous designs incorporated cooling water pipes embedded in the concrete shielding wall for thermal management. However, we determined that this approach failed to maintain concrete temperatures within specified limits due to the low thermal conductivity of concrete and the challenge of ensuring adequate contact between the concrete and piping. We developed a new shielding wall design that eliminates the need for cooling pipes inside the concrete. This updated design integrates an 85 cm of stainless steel 316 L (SS316L) structure containing 20 % cooling water positioned in front of the concrete, which maintains concrete temperatures within acceptable limits. In addition, nuclear assessments of this revised shielding wall structure were conducted to verify its radiation shielding capacity. Results indicate that the effective dose criteria are met at thicknesses of 255 cm for heavy concrete and 335 cm for ordinary concrete when the 85 cm of combing structure of SS316L/water (20 % water) is included.
{"title":"Nuclear, thermal, and shielding design of test cell in A-FNS","authors":"Shogo Honda, Saerom Kwon, Shunsuke Kenjo, Makoto Oyaidzu, Kentaro Ochiai, Satoshi Sato","doi":"10.1016/j.fusengdes.2025.115549","DOIUrl":"10.1016/j.fusengdes.2025.115549","url":null,"abstract":"<div><div>Significant quantities of neutrons and gamma rays deduce the nuclear heating on the shielding walls of the test cell (TC) in the A-FNS. This study presents nuclear and thermal analyses of the shielding wall in the A-FNS to evaluate its radiation-shielding effectiveness and cooling capacity. Previous designs incorporated cooling water pipes embedded in the concrete shielding wall for thermal management. However, we determined that this approach failed to maintain concrete temperatures within specified limits due to the low thermal conductivity of concrete and the challenge of ensuring adequate contact between the concrete and piping. We developed a new shielding wall design that eliminates the need for cooling pipes inside the concrete. This updated design integrates an 85 cm of stainless steel 316 L (SS316L) structure containing 20 % cooling water positioned in front of the concrete, which maintains concrete temperatures within acceptable limits. In addition, nuclear assessments of this revised shielding wall structure were conducted to verify its radiation shielding capacity. Results indicate that the effective dose criteria are met at thicknesses of 255 cm for heavy concrete and 335 cm for ordinary concrete when the 85 cm of combing structure of SS316L/water (20 % water) is included.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115549"},"PeriodicalIF":2.0,"publicationDate":"2025-12-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694584","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.2025.115551
C.S. Chisholm , T. Berry, D.T. Garnier, R.A. Badcock, G. Bioletti, K. Bouloukakis , E. Brewerton, M.A. Buchanan, P.J. Burt, E.V.W. Chambers, K.B. Chappell, P. Coulson, R.J. Davidson, J.P.M. Ellingham, P. Geursen, K. Hamilton, R. Hu, E. Hunter, J.P. Jones, P. Kusay, N. Zhou
OpenStar Technologies is a private fusion company exploring the levitated dipole concept for commercial fusion energy production. OpenStar has manufactured a new generation of levitated dipole experiment, called “Junior”, leveraging recent advances made in high-temperature superconducting magnet technologies. Junior houses a 5.6 T REBCO high-temperature superconducting magnet in a 5.2 m vacuum chamber, with plasma heating achieved via 50 kW of electron cyclotron resonance heating power. Importantly, this experiment integrates novel high temperature superconductor power supply technology on board the dipole magnet. Recently OpenStar has completed first experimental campaigns with the Junior experiment, achieving first plasmas in late 2024. Experiments conducted with the full levitated system are planned for 2025. This article provides an overview of the main results from these experiments and details improvements planned for future campaigns.
{"title":"Design and initial results from the “Junior” Levitated Dipole Experiment","authors":"C.S. Chisholm , T. Berry, D.T. Garnier, R.A. Badcock, G. Bioletti, K. Bouloukakis , E. Brewerton, M.A. Buchanan, P.J. Burt, E.V.W. Chambers, K.B. Chappell, P. Coulson, R.J. Davidson, J.P.M. Ellingham, P. Geursen, K. Hamilton, R. Hu, E. Hunter, J.P. Jones, P. Kusay, N. Zhou","doi":"10.1016/j.fusengdes.2025.115551","DOIUrl":"10.1016/j.fusengdes.2025.115551","url":null,"abstract":"<div><div>OpenStar Technologies is a private fusion company exploring the levitated dipole concept for commercial fusion energy production. OpenStar has manufactured a new generation of levitated dipole experiment, called “Junior”, leveraging recent advances made in high-temperature superconducting magnet technologies. Junior houses a <span><math><mo>∼</mo></math></span>5.6 T REBCO high-temperature superconducting magnet in a 5.2 m vacuum chamber, with plasma heating achieved via <span><math><mo><</mo></math></span>50 kW of electron cyclotron resonance heating power. Importantly, this experiment integrates novel high temperature superconductor power supply technology on board the dipole magnet. Recently OpenStar has completed first experimental campaigns with the Junior experiment, achieving first plasmas in late 2024. Experiments conducted with the full levitated system are planned for 2025. This article provides an overview of the main results from these experiments and details improvements planned for future campaigns.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"223 ","pages":"Article 115551"},"PeriodicalIF":2.0,"publicationDate":"2025-12-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145694772","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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