Pub Date : 2026-02-01Epub 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":"2026-02-01","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 : 2026-01-01Epub Date: 2025-09-19DOI: 10.1016/j.fusengdes.2025.115450
Wanjing Wang , Lingming Huang , Jichao Wang , Zhenjie Zhang , Qun Li , Zhenmao Chen , Peisong Du , Haishan Zhou , Guang-Nan Luo
The double wall tubes (DWT) in the tritium breeding zone play a critical role in the Water Coolant Ceramics Blanket (WCCB) of the China Fusion Engineering Testing Reactor (CFETR). In order to assess the effectiveness of DWT in resisting type-I crack propagation, a sandwich structure plate consisting of steel-interlayer-steel was fabricated using heat isostatic press (HIP) technology. Subsequently, experimental investigations were carried out to study fatigue crack propagation under three-point bending conditions. The results demonstrate that continuous fatigue bending leads to the generation of a type-I crack perpendicular to the interface; however, there is stagnation at the interface before expanding sideways to form a type-II crack. Furthermore, it was observed that specimens with Ni interlayers exhibit more effective resistance against type-II crack extension compared to those with Cu interlayers. In the discussion, a J-integral-based fatigue crack propagation model was proposed, which can accurately predict the deflection behavior of cracks in multilayer structures with interfaces. This study indicates that the DWT could effectively prevent the propagation of type-I cracks.
{"title":"Effects of interlayer on fracture and fatigue crack resisting of double-wall tubes in WCCB","authors":"Wanjing Wang , Lingming Huang , Jichao Wang , Zhenjie Zhang , Qun Li , Zhenmao Chen , Peisong Du , Haishan Zhou , Guang-Nan Luo","doi":"10.1016/j.fusengdes.2025.115450","DOIUrl":"10.1016/j.fusengdes.2025.115450","url":null,"abstract":"<div><div>The double wall tubes (DWT) in the tritium breeding zone play a critical role in the Water Coolant Ceramics Blanket (WCCB) of the China Fusion Engineering Testing Reactor (CFETR). In order to assess the effectiveness of DWT in resisting type-I crack propagation, a sandwich structure plate consisting of steel-interlayer-steel was fabricated using heat isostatic press (HIP) technology. Subsequently, experimental investigations were carried out to study fatigue crack propagation under three-point bending conditions. The results demonstrate that continuous fatigue bending leads to the generation of a type-I crack perpendicular to the interface; however, there is stagnation at the interface before expanding sideways to form a type-II crack. Furthermore, it was observed that specimens with Ni interlayers exhibit more effective resistance against type-II crack extension compared to those with Cu interlayers. In the discussion, a J-integral-based fatigue crack propagation model was proposed, which can accurately predict the deflection behavior of cracks in multilayer structures with interfaces. This study indicates that the DWT could effectively prevent the propagation of type-I cracks.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115450"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145096618","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 : 2026-01-01Epub Date: 2025-11-24DOI: 10.1016/j.fusengdes.2025.115533
Wei Zheng , Qian Xu , Jichan Xu , Xin Yang , Haishan Zhou , Guangnan Luo
To meet the demanding requirements for high-precision, spatially-resolved diagnostics in plasma-material interaction (PMI) studies under the high-flux, high-magnetic-field environment of the SPARROW linear plasma device, an actively water-cooled Langmuir probe array system has been designed and developed. This design synergizes actively water-cooling with array layout requirements, with a focus on optimizing the cooling channel structure. Through systematic computational fluid dynamics (CFD) simulations, the thermal performance of the probe was quantitatively evaluated under Gaussian-distributed heat fluxes of 10 MW/m², 15 MW/m², and 20 MW/m², along with the impact on the probe body and the key insulating material (alumina ceramic). Under the 15 MW/m² heat flux, the maximum temperatures of the tungsten tip and alumina sleeve are maintained at approximately 62% and 61% of their respective safety limits. Even under the extreme 20 MW/m² condition, these key diagnostic components remain below 75% of their limits, demonstrating a substantial safety buffer that accommodates potential CFD uncertainties. By integrating innovative design with comprehensive thermal analysis, this research establishes key technical foundations for achieving efficient and reliable arrayed active diagnostics in extreme fusion-relevant plasma environments. It provides vital support for future high-parameter plasma physics experiments.
{"title":"Design and thermal analysis of an actively water-cooled array probe for the SPARROW device","authors":"Wei Zheng , Qian Xu , Jichan Xu , Xin Yang , Haishan Zhou , Guangnan Luo","doi":"10.1016/j.fusengdes.2025.115533","DOIUrl":"10.1016/j.fusengdes.2025.115533","url":null,"abstract":"<div><div>To meet the demanding requirements for high-precision, spatially-resolved diagnostics in plasma-material interaction (PMI) studies under the high-flux, high-magnetic-field environment of the SPARROW linear plasma device, an actively water-cooled Langmuir probe array system has been designed and developed. This design synergizes actively water-cooling with array layout requirements, with a focus on optimizing the cooling channel structure. Through systematic computational fluid dynamics (CFD) simulations, the thermal performance of the probe was quantitatively evaluated under Gaussian-distributed heat fluxes of 10 MW/m², 15 MW/m², and 20 MW/m², along with the impact on the probe body and the key insulating material (alumina ceramic). Under the 15 MW/m² heat flux, the maximum temperatures of the tungsten tip and alumina sleeve are maintained at approximately 62% and 61% of their respective safety limits. Even under the extreme 20 MW/m² condition, these key diagnostic components remain below 75% of their limits, demonstrating a substantial safety buffer that accommodates potential CFD uncertainties. By integrating innovative design with comprehensive thermal analysis, this research establishes key technical foundations for achieving efficient and reliable arrayed active diagnostics in extreme fusion-relevant plasma environments. It provides vital support for future high-parameter plasma physics experiments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115533"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684658","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 : 2026-01-01Epub Date: 2025-09-30DOI: 10.1016/j.fusengdes.2025.115454
Fabrizio Lisanti , Alex Aimetta , Pietro Arena , Roberto Bonifetto , Antonio Froio
One of the main milestones towards the development of the EU DEMO reactor is to demonstrate the feasibility of a closed tritium fuel cycle, a key aspect for the generation of electricity from fusion energy by the middle of the century. In view of this, the design of the breeding blanket (BB) has a key role. A candidate design for the EU DEMO BB is the Water-Cooled Lithium-Lead (WCLL) concept, where eutectic lithium-lead (PbLi) is circulated in a suitable closed circuit. A key issue in the design of the PbLi circuit is the evaluation of the inventories of Activated Corrosion Products (ACPs), which are solid particles corroded from structural materials and eventually activated in the blanket, transported inside the loop within the PbLi. In recent years, a PbLi loop model has been implemented in the GETTHEM code, a system-level tool for the thermal-hydraulic modelling of BB and related subsystems. In this work, in addition to the already existing assessment of corrosion phenomena, models of different pieces of physics necessary for a comprehensive assessment of the ACP inventories are added to the PbLi loop model in GETTHEM. Specifically, these include activation and decay of the corroded species in the BB. For the latter, a sink term for the radioactive decay and a source term for the transmutation due to neutrons interaction with materials are introduced in the mass conservation equations for each ACP. To demonstrate the code capabilities, a representative test case is presented.
{"title":"Multiphysics modelling of the Activated Corrosion Products generation and transport in the WCLL PbLi loop with GETTHEM","authors":"Fabrizio Lisanti , Alex Aimetta , Pietro Arena , Roberto Bonifetto , Antonio Froio","doi":"10.1016/j.fusengdes.2025.115454","DOIUrl":"10.1016/j.fusengdes.2025.115454","url":null,"abstract":"<div><div>One of the main milestones towards the development of the EU DEMO reactor is to demonstrate the feasibility of a closed tritium fuel cycle, a key aspect for the generation of electricity from fusion energy by the middle of the century. In view of this, the design of the breeding blanket (BB) has a key role. A candidate design for the EU DEMO BB is the Water-Cooled Lithium-Lead (WCLL) concept, where eutectic lithium-lead (PbLi) is circulated in a suitable closed circuit. A key issue in the design of the PbLi circuit is the evaluation of the inventories of Activated Corrosion Products (ACPs), which are solid particles corroded from structural materials and eventually activated in the blanket, transported inside the loop within the PbLi. In recent years, a PbLi loop model has been implemented in the GETTHEM code, a system-level tool for the thermal-hydraulic modelling of BB and related subsystems. In this work, in addition to the already existing assessment of corrosion phenomena, models of different pieces of physics necessary for a comprehensive assessment of the ACP inventories are added to the PbLi loop model in GETTHEM. Specifically, these include activation and decay of the corroded species in the BB. For the latter, a sink term for the radioactive decay and a source term for the transmutation due to neutrons interaction with materials are introduced in the mass conservation equations for each ACP. To demonstrate the code capabilities, a representative test case is presented.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115454"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221038","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-01Epub Date: 2025-10-14DOI: 10.1016/j.fusengdes.2025.115451
Dhinesh Thanganadar, Jack Acres
The Spherical Tokamak for Energy Production (STEP) programme, led by UK Industrial Fusion Solutions Ltd (UKIFS), aims to build a prototype fusion power plant by the 2040s, paving the way for commercial fusion energy. A key challenge is efficiently converting fusion power into net electrical power, given the conflicting requirements of high power conversion efficiency and integration of multiple heat grades from various tokamak components. Additionally, the power cycle must exhibit high operational flexibility and reliability for intermittent pulse mode operation during initial phases. This work addresses the research gap by evaluating several power cycle concepts for the STEP prototype, resulting in unique designs. Novel power cycles have been developed and optimised based on steady-state performance, considering efficiency, commercial viability, and safety. Three power cycle designs are presented: 1) steam Rankine cycle, 2) hybrid steam-organic Rankine cycle, and 3) supercritical CO2 Brayton cycle. These configurations have been modelled and evaluated to compare design performance parameters such as cycle efficiency, net power, technology readiness, and safety aspects, aiding in the selection of technologies for fusion power cycle.
{"title":"Thermodynamic performance evaluation of power cycle technologies for spherical Tokamak Energy Production","authors":"Dhinesh Thanganadar, Jack Acres","doi":"10.1016/j.fusengdes.2025.115451","DOIUrl":"10.1016/j.fusengdes.2025.115451","url":null,"abstract":"<div><div>The Spherical Tokamak for Energy Production (STEP) programme, led by UK Industrial Fusion Solutions Ltd (UKIFS), aims to build a prototype fusion power plant by the 2040s, paving the way for commercial fusion energy. A key challenge is efficiently converting fusion power into net electrical power, given the conflicting requirements of high power conversion efficiency and integration of multiple heat grades from various tokamak components. Additionally, the power cycle must exhibit high operational flexibility and reliability for intermittent pulse mode operation during initial phases. This work addresses the research gap by evaluating several power cycle concepts for the STEP prototype, resulting in unique designs. Novel power cycles have been developed and optimised based on steady-state performance, considering efficiency, commercial viability, and safety. Three power cycle designs are presented: 1) steam Rankine cycle, 2) hybrid steam-organic Rankine cycle, and 3) supercritical CO<sub>2</sub> Brayton cycle. These configurations have been modelled and evaluated to compare design performance parameters such as cycle efficiency, net power, technology readiness, and safety aspects, aiding in the selection of technologies for fusion power cycle.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115451"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145320569","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 : 2026-01-01Epub Date: 2025-10-14DOI: 10.1016/j.fusengdes.2025.115493
Diogo Rechena , João da Silva , Conceição Amado , Virgínia Infante , Paulo Varela , Jorge Manuel Santos , António Silva , Bruno Gonçalves , Liu Yong , Victor Udintsev
ITER will explore nuclear fusion under unprecedented operational conditions in terms of neutron exposure, magnetic fields and fusion power. Diagnostics play a vital role in the operation of ITER, providing plasma control, machine protection, and allowing for physics experiments, whilst withstanding these extreme conditions. The ITER Reliability, Availability, Maintainability, and Inspectability (RAMI) program was developed to address the impact of these operational conditions on its systems from the early design phases. However, the specifics of each diagnostic operation are not considered when determining availability requirements. In this work, we present simulations of the Reliability Block Diagram (RBD) of the ITER Collective Thomson Scattering (CTS) diagnostic to obtain its mean inherent availability distribution, using the methodology to estimate the availability distribution parameters to compare the effects of different operation and maintenance conditions. Finally, we assess the applicability of the mean inherent availability as a system design requirement and show that it is only a suitable metric for systems that can undergo maintenance at any time. Additionally, we show that diagnostics need alternative requirements that consider the spread of the mean availability distribution. For that purpose, we propose two types of requirements for the design of diagnostics: the first uses the mean availability and its standard deviation, while the second sets probability limits for desirable and undesirable availability values.
{"title":"Availability requirements for diagnostics in nuclear fusion: the ITER Collective Thomson Scattering case-study","authors":"Diogo Rechena , João da Silva , Conceição Amado , Virgínia Infante , Paulo Varela , Jorge Manuel Santos , António Silva , Bruno Gonçalves , Liu Yong , Victor Udintsev","doi":"10.1016/j.fusengdes.2025.115493","DOIUrl":"10.1016/j.fusengdes.2025.115493","url":null,"abstract":"<div><div>ITER will explore nuclear fusion under unprecedented operational conditions in terms of neutron exposure, magnetic fields and fusion power. Diagnostics play a vital role in the operation of ITER, providing plasma control, machine protection, and allowing for physics experiments, whilst withstanding these extreme conditions. The ITER Reliability, Availability, Maintainability, and Inspectability (RAMI) program was developed to address the impact of these operational conditions on its systems from the early design phases. However, the specifics of each diagnostic operation are not considered when determining availability requirements. In this work, we present simulations of the Reliability Block Diagram (RBD) of the ITER Collective Thomson Scattering (CTS) diagnostic to obtain its mean inherent availability distribution, using the methodology to estimate the availability distribution parameters to compare the effects of different operation and maintenance conditions. Finally, we assess the applicability of the mean inherent availability as a system design requirement and show that it is only a suitable metric for systems that can undergo maintenance at any time. Additionally, we show that diagnostics need alternative requirements that consider the spread of the mean availability distribution. For that purpose, we propose two types of requirements for the design of diagnostics: the first uses the mean availability and its standard deviation, while the second sets probability limits for desirable and undesirable availability values.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115493"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145320568","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}
The ADITYA tokamak (R₀ = 0.75 m, a = 0.25 m), originally operated with a limiter configuration, has been successfully upgraded to ADITYA-U with an open divertor configuration to enhance plasma confinement and operational flexibility. Limiter and divertor systems define the plasma boundary within the vacuum vessel, protecting in-vessel components by minimizing direct plasma-wall interactions. In both single-null and double-null divertor configurations, ADITYA-U is designed to produce circular and shaped plasmas with a triangularity (δ) of ∼0.45, elongation (κ) of ∼1.1–1.2, and plasma currents (Ip) in the range of 100–150 kA.
Using the plasma equilibrium simulation code IPREQ, the optimal locations for limiter and divertor plates were determined and validated for the new vacuum vessel geometry. Plasma-facing components (PFCs) based on graphite—including toroidal, poloidal, and safety limiters, as well as divertor tiles—were installed in a staged manner to facilitate a progressive operational strategy. Limiters were installed prior to the initial operation phase to manage plasma-wall interactions during early campaigns, while divertor plates were added subsequently, following operational experience with impurity behavior during the burn-through phase.
This paper details the preparatory studies, simulation-guided design process, and the in-situ installation challenges associated with retrofitting graphite limiter and divertor assemblies in the ADITYA-U tokamak. The work provides insights into the phased upgrade process of a medium-sized tokamak and highlights practical strategies for integrating advanced plasma boundary configurations in existing devices.
ADITYA托卡马克（R 0 = 0.75 m, a = 0.25 m）最初采用限流器配置，现已成功升级为ADITYA- u，采用开放式导流器配置，以增强等离子体约束和操作灵活性。限制器和分流器系统定义了真空容器内的等离子体边界，通过最大限度地减少等离子体与容器壁的直接相互作用来保护容器内的组件。在单零和双零分流器配置中，ADITYA-U设计用于产生三角形（δ）为~ 0.45，伸长率（κ）为~ 1.1-1.2的圆形和形状等离子体，等离子体电流（Ip）在100-150 kA范围内。利用等离子体平衡模拟代码IPREQ，确定了限制板和导流板的最佳位置，并对新真空容器的几何形状进行了验证。基于石墨的等离子体组件（pfc），包括环向、极向、安全限制器以及导流砖，都是分阶段安装的，以促进渐进式作业策略。在初始操作阶段之前安装了限位器，以管理早期活动期间等离子体壁的相互作用，随后根据在烧透阶段的杂质行为的操作经验添加了导流板。本文详细介绍了ADITYA-U托卡马克中石墨限位器和导流器组件改造的前期研究、仿真指导设计过程以及现场安装挑战。这项工作为中型托卡马克的分阶段升级过程提供了见解，并强调了在现有设备中集成先进等离子体边界配置的实用策略。
{"title":"Design, construction, integration and installation of plasma facing components of limiter & divertor of ADITYA-U tokamak","authors":"K.M. Patel , K.A. Jadeja , Harshita Raj , J. Ghosh , S.B. Bhatt , R.L. Tanna , Deepti Sharma , Arun Prakash , Rupesh G , Suman Aich , Komal Yadav , SK Injamul Hoque , Rohit Kumar , Aditya-U Team","doi":"10.1016/j.fusengdes.2025.115520","DOIUrl":"10.1016/j.fusengdes.2025.115520","url":null,"abstract":"<div><div>The ADITYA tokamak (R₀ = 0.75 m, <em>a</em> = 0.25 m), originally operated with a limiter configuration, has been successfully upgraded to ADITYA-U with an open divertor configuration to enhance plasma confinement and operational flexibility. Limiter and divertor systems define the plasma boundary within the vacuum vessel, protecting in-vessel components by minimizing direct plasma-wall interactions. In both single-null and double-null divertor configurations, ADITYA-U is designed to produce circular and shaped plasmas with a triangularity (δ) of ∼0.45, elongation (κ) of ∼1.1–1.2, and plasma currents (Ip) in the range of 100–150 kA.</div><div>Using the plasma equilibrium simulation code IPREQ, the optimal locations for limiter and divertor plates were determined and validated for the new vacuum vessel geometry. Plasma-facing components (PFCs) based on graphite—including toroidal, poloidal, and safety limiters, as well as divertor tiles—were installed in a staged manner to facilitate a progressive operational strategy. Limiters were installed prior to the initial operation phase to manage plasma-wall interactions during early campaigns, while divertor plates were added subsequently, following operational experience with impurity behavior during the burn-through phase.</div><div>This paper details the preparatory studies, simulation-guided design process, and the in-situ installation challenges associated with retrofitting graphite limiter and divertor assemblies in the ADITYA-U tokamak. The work provides insights into the phased upgrade process of a medium-sized tokamak and highlights practical strategies for integrating advanced plasma boundary configurations in existing devices.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115520"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145519558","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 : 2026-01-01Epub Date: 2025-11-05DOI: 10.1016/j.fusengdes.2025.115510
F. Stelzer , S. An , T. Hohmann , V. Rohde , ASDEX Upgrade Team
High performance operation of magnetic confinement fusion reactors requires conditioning of the plasma-facing components by boronization [1]. This boron hydride layers traps residual oxygen in the vacuum vessel, thereby reducing radiation losses and, especially in metal components, mitigating sputtering by oxygen ions. Technically, at ASDEX Upgrade this coating is formed by a glow discharge process, using 10 % deuterated diborane (B2D6) in helium as carrier gas (hereafter referred to as the “B₂D₆/He mix”). In this publication, we report on the complex software architecture, which was generally realized in the form of finite state machines (FSM). These state machines contain the entire process control of the boronization process, incorporating all the safety interlocks stored in a locking matrix. With this new procedure, the boronization can be carried out within 6 h on a normal working day, without the need to evacuate the office buildings adjacent to the experimental hall.
{"title":"Efficient boronization through automation: A case study of Simatic software at ASDEX upgrade","authors":"F. Stelzer , S. An , T. Hohmann , V. Rohde , ASDEX Upgrade Team","doi":"10.1016/j.fusengdes.2025.115510","DOIUrl":"10.1016/j.fusengdes.2025.115510","url":null,"abstract":"<div><div>High performance operation of magnetic confinement fusion reactors requires conditioning of the plasma-facing components by boronization [<span><span>1]</span></span>. This boron hydride layers traps residual oxygen in the vacuum vessel, thereby reducing radiation losses and, especially in metal components, mitigating sputtering by oxygen ions. Technically, at ASDEX Upgrade this coating is formed by a glow discharge process, using 10 % deuterated diborane (B<sub>2</sub>D<sub>6</sub>) in helium as carrier gas (hereafter referred to as the “B₂D₆/He mix”). In this publication, we report on the complex software architecture, which was generally realized in the form of finite state machines (FSM). These state machines contain the entire process control of the boronization process, incorporating all the safety interlocks stored in a locking matrix. With this new procedure, the boronization can be carried out within 6 h on a normal working day, without the need to evacuate the office buildings adjacent to the experimental hall.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115510"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145465524","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 : 2026-01-01Epub Date: 2025-11-04DOI: 10.1016/j.fusengdes.2025.115507
Linxian Che , Shikun Wen , Xin Huang , Guodong Qin , Shijie Liu
This paper proposes a hybrid series-parallel assembly manipulator system in response to the remote handling task requirements for auxiliary assembly and maintenance of vacuum vessels in nuclear fusion. The proposed system incorporates an annular moving rail, transport platform, lifting bracket, and a novel 2RUPaR-2RSS parallel assembled manipulator (PAM) with three translational and one rotational (3T1R) degrees of freedom (DOFs). The PAM structure features a symmetrical design that avoids over-constraints, and its modular installation on transport rails enables efficient in-situ maintenance operations, encompassing vacuum vessel assembly, precision cutting, arc welding, and other associated processes. This study employs screw theory to analyze the constraint characteristics and DOFs properties of PAM. Displacement analysis equations for the PAM are formulated based on link length constraints. Analytical expressions for velocity and acceleration are derived, followed by an investigation into the singular configurations of the PAM through its velocity Jacobian matrix. Based on analyzing transmission indices for the branch chain in the PAM, this work defines the effective transmission workspace and the global transmission indices depending on the permissible transmission indices. It demonstrates an analytical example for the mechanism workspace. This study can provide theoretical foundations for applying the novel PAM in fusion plants.
{"title":"Kinematics performance analysis of a 2RUPaR-2RSS parallel assembled manipulator for vacuum vessel in fusion plants","authors":"Linxian Che , Shikun Wen , Xin Huang , Guodong Qin , Shijie Liu","doi":"10.1016/j.fusengdes.2025.115507","DOIUrl":"10.1016/j.fusengdes.2025.115507","url":null,"abstract":"<div><div>This paper proposes a hybrid series-parallel assembly manipulator system in response to the remote handling task requirements for auxiliary assembly and maintenance of vacuum vessels in nuclear fusion. The proposed system incorporates an annular moving rail, transport platform, lifting bracket, and a novel 2RUP<sub>a</sub>R-2RSS parallel assembled manipulator (PAM) with three translational and one rotational (3T1R) degrees of freedom (DOFs). The PAM structure features a symmetrical design that avoids over-constraints, and its modular installation on transport rails enables efficient in-situ maintenance operations, encompassing vacuum vessel assembly, precision cutting, arc welding, and other associated processes. This study employs screw theory to analyze the constraint characteristics and DOFs properties of PAM. Displacement analysis equations for the PAM are formulated based on link length constraints. Analytical expressions for velocity and acceleration are derived, followed by an investigation into the singular configurations of the PAM through its velocity Jacobian matrix. Based on analyzing transmission indices for the branch chain in the PAM, this work defines the effective transmission workspace and the global transmission indices depending on the permissible transmission indices. It demonstrates an analytical example for the mechanism workspace. This study can provide theoretical foundations for applying the novel PAM in fusion plants.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115507"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145465523","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 : 2026-01-01Epub Date: 2025-10-20DOI: 10.1016/j.fusengdes.2025.115495
Liming Zhang , Peng Liu , Siqing Feng , Zhihong Liu , Xuebing Peng
The divertor, a key component in tokamak fusion devices, currently features two mainstream target designs: flat-type and monoblock-type. Achieving the flat-type configuration necessitates a reliable joint between CuCrZr and SS316L. Explosive welding is identified as a suitable method for fabricating the CuCrZr/SS316L composite plate, which is critical for ensuring robust joint strength. The high temperature heat treatment applied after explosive welding has a significant influence on the strength of the CuCrZr/SS316L joint. This study examines the performance of CuCrZr/SS316L joint subjected to various heat treatment schemes, including conventional stress relief, vacuum brazing, solution annealing, and aging treatment. Tensile testing, hardness measurements, and metallographic analysis were employed to assess the strength of the welded joints under different thermal conditions. The findings demonstrate that aging treatment effectively counteracts the performance degradation of CuCrZr induced by solution annealing, leading to a significant improvement in the overall performance of the welded joint.
{"title":"Experimental studies for CuCrZr/SS316L joint of plasma facing component in flat-type divertor","authors":"Liming Zhang , Peng Liu , Siqing Feng , Zhihong Liu , Xuebing Peng","doi":"10.1016/j.fusengdes.2025.115495","DOIUrl":"10.1016/j.fusengdes.2025.115495","url":null,"abstract":"<div><div>The divertor, a key component in tokamak fusion devices, currently features two mainstream target designs: flat-type and monoblock-type. Achieving the flat-type configuration necessitates a reliable joint between CuCrZr and SS316L. Explosive welding is identified as a suitable method for fabricating the CuCrZr/SS316L composite plate, which is critical for ensuring robust joint strength. The high temperature heat treatment applied after explosive welding has a significant influence on the strength of the CuCrZr/SS316L joint. This study examines the performance of CuCrZr/SS316L joint subjected to various heat treatment schemes, including conventional stress relief, vacuum brazing, solution annealing, and aging treatment. Tensile testing, hardness measurements, and metallographic analysis were employed to assess the strength of the welded joints under different thermal conditions. The findings demonstrate that aging treatment effectively counteracts the performance degradation of CuCrZr induced by solution annealing, leading to a significant improvement in the overall performance of the welded joint.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115495"},"PeriodicalIF":2.0,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145362504","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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