Pub Date : 2024-12-01DOI: 10.1016/j.fusengdes.2024.114739
Louis Butt , Alexandra Dickinson-Lomas , Martin Freer , Joven Lim , Yu-Lung Chiu
The status of research and development on vanadium alloys and on their application as a structural material in the breeder blanket of a fusion power plant are reviewed. Emphasis has been placed on irradiation experiments and the effects of neutron and ion irradiation on the microstructure of the current reference alloy, V4Cr4Ti. The effects of a liquid lithium environment and the exchange of carbon, nitrogen, and oxygen impurities between the alloy and application-relevant environments are also reviewed. The dependence of microstructural features on irradiation experiment parameters are examined, with a summary on magnetohydrodynamic effects and mechanical properties of the alloy. The relationship between the ductility and the alloy chemistry and the irradiation test temperature are analysed. Review of the state of the art indicates that further developments of facilities simulating a realistic fusion radiation and liquid lithium environment are fundamental to advancing fusion reactor breeder design.
{"title":"Research and development on vanadium alloys for fusion breeder blanket application","authors":"Louis Butt , Alexandra Dickinson-Lomas , Martin Freer , Joven Lim , Yu-Lung Chiu","doi":"10.1016/j.fusengdes.2024.114739","DOIUrl":"10.1016/j.fusengdes.2024.114739","url":null,"abstract":"<div><div>The status of research and development on vanadium alloys and on their application as a structural material in the breeder blanket of a fusion power plant are reviewed. Emphasis has been placed on irradiation experiments and the effects of neutron and ion irradiation on the microstructure of the current reference alloy, V<img>4Cr<img>4Ti. The effects of a liquid lithium environment and the exchange of carbon, nitrogen, and oxygen impurities between the alloy and application-relevant environments are also reviewed. The dependence of microstructural features on irradiation experiment parameters are examined, with a summary on magnetohydrodynamic effects and mechanical properties of the alloy. The relationship between the ductility and the alloy chemistry and the irradiation test temperature are analysed. Review of the state of the art indicates that further developments of facilities simulating a realistic fusion radiation and liquid lithium environment are fundamental to advancing fusion reactor breeder design.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114739"},"PeriodicalIF":1.9,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142756896","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-29DOI: 10.1016/j.fusengdes.2024.114737
John Waterhouse, Martin Wheatley, Adam Stephen, Colin Hogben, Graham Jones, Alex Goodyear, Tom Farmer, Paul McCullen
The JET Control and Data Acquisition System (CODAS) has stood the test of time and seen us through to the end of JET plasma operations in 2023. The system architecture has remained largely un-changed over the last decade or so although many new diagnostics and control systems have been added and the volume of data collected has grown massively. CAMAC remains at the heart of the system, particularly for continuous acquisition and control for much of the traditional, stable parts of the system. However, most of the newer diagnostics and control systems are network attached. The CAMAC interface was changed, about a decade ago, to remove the serial highway driver from the subsystem host so that the subsystem hosts could be virtualised and run on more powerful Oracle/Sun hosts, with the Serial Highway driver running on legacy sub-system host network attached. Other significant changes have been the development of a standard, web-based interface for control and data acquisition for diagnostics, the adoption of EPICS for several diagnostics and plant control along with integration into CODAS and the adoption of the ITER SDN protocols over ethernet to supplement the original ATM based real time control infrastructure.
These developments were driven by the increased monitoring and machine protection requirements for the ITER Like Wall and the enhanced requirements for the high power Tritium Campaigns (DTE2 and DTE3). The latter, including significant expansion of the neutron and gamma diagnostics, along with expansion of the Tritium introduction systems and enhanced control systems. Towards the end of plasma operations, the requirement for the Laser Induced Desorption with Quadrupole Mass Spectroscopy (LIDS-QMS) involved significant development work to incorporate the associated control and data acquisition systems together with pushing the mode of operation for JET Pulses. At the very end of plasma operations, the requirements for long pulse operation also pushed the pulse operating mode. We now progress to Decommissioning and Repurposing JET and CODAS continues to be adapted to support the diminishing number of systems required to support the plant that is still operational and support diagnostic calibration later this year.
{"title":"JET CODAS - the final status","authors":"John Waterhouse, Martin Wheatley, Adam Stephen, Colin Hogben, Graham Jones, Alex Goodyear, Tom Farmer, Paul McCullen","doi":"10.1016/j.fusengdes.2024.114737","DOIUrl":"10.1016/j.fusengdes.2024.114737","url":null,"abstract":"<div><div>The JET Control and Data Acquisition System (CODAS) has stood the test of time and seen us through to the end of JET plasma operations in 2023. The system architecture has remained largely un-changed over the last decade or so although many new diagnostics and control systems have been added and the volume of data collected has grown massively. CAMAC remains at the heart of the system, particularly for continuous acquisition and control for much of the traditional, stable parts of the system. However, most of the newer diagnostics and control systems are network attached. The CAMAC interface was changed, about a decade ago, to remove the serial highway driver from the subsystem host so that the subsystem hosts could be virtualised and run on more powerful Oracle/Sun hosts, with the Serial Highway driver running on legacy sub-system host network attached. Other significant changes have been the development of a standard, web-based interface for control and data acquisition for diagnostics, the adoption of EPICS for several diagnostics and plant control along with integration into CODAS and the adoption of the ITER SDN protocols over ethernet to supplement the original ATM based real time control infrastructure.</div><div>These developments were driven by the increased monitoring and machine protection requirements for the ITER Like Wall and the enhanced requirements for the high power Tritium Campaigns (DTE2 and DTE3). The latter, including significant expansion of the neutron and gamma diagnostics, along with expansion of the Tritium introduction systems and enhanced control systems. Towards the end of plasma operations, the requirement for the Laser Induced Desorption with Quadrupole Mass Spectroscopy (LIDS-QMS) involved significant development work to incorporate the associated control and data acquisition systems together with pushing the mode of operation for JET Pulses. At the very end of plasma operations, the requirements for long pulse operation also pushed the pulse operating mode. We now progress to Decommissioning and Repurposing JET and CODAS continues to be adapted to support the diminishing number of systems required to support the plant that is still operational and support diagnostic calibration later this year.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114737"},"PeriodicalIF":1.9,"publicationDate":"2024-11-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142745623","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper presents design considerations of a new power supply and a new coil system for the vertical plasma position control in ADITYA-U tokamak. The main objective is to improve overall plasma performance compared to the existing open-loop system. Feedback stabilization of the plasma vertical position in ADITYA-U tokamak will play a vital role. Vertical Displacement Events (VDE) can be triggered during shaped plasma operation and requires proper control to avoid plasma disruptions. A radial magnetic field created by exterior coils will arrest the vertical plasma position against unwanted VDEs. Therefore, procurement of a new power supply namely, active position magnet control power supply (APMC-PS) is proposed to fulfill the vertical position control requirement in the ADITYA-U tokamak. This work presents the overall requirement for an active control system to improve the vertical control capability in the mid-size machine without passive stabilisers. The proposed scheme will support a reduction in time response with relatively less apparent power for controlling the vertical plasma position.
{"title":"Study and analysis of the design considerations for controlling vertical plasma position in ADITYA-U tokamak","authors":"Rohit Kumar , Harshita Raj , Vinay Menon , Deepti Sharma , Darshan Parmar , Dinesh Sharma , Vishal Jain , YSS Srinivas , Shivam Gupta , Rakesh Tanna , Ashok Mankani , Joydeep Ghosh","doi":"10.1016/j.fusengdes.2024.114736","DOIUrl":"10.1016/j.fusengdes.2024.114736","url":null,"abstract":"<div><div>This paper presents design considerations of a new power supply and a new coil system for the vertical plasma position control in ADITYA-U tokamak. The main objective is to improve overall plasma performance compared to the existing open-loop system. Feedback stabilization of the plasma vertical position in ADITYA-U tokamak will play a vital role. Vertical Displacement Events (VDE) can be triggered during shaped plasma operation and requires proper control to avoid plasma disruptions. A radial magnetic field created by exterior coils will arrest the vertical plasma position against unwanted VDEs. Therefore, procurement of a new power supply namely, active position magnet control power supply (APMC-PS) is proposed to fulfill the vertical position control requirement in the ADITYA-U tokamak. This work presents the overall requirement for an active control system to improve the vertical control capability in the mid-size machine without passive stabilisers. The proposed scheme will support a reduction in time response with relatively less apparent power for controlling the vertical plasma position.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114736"},"PeriodicalIF":1.9,"publicationDate":"2024-11-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142745624","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 : 2024-11-26DOI: 10.1016/j.fusengdes.2024.114732
M.D. Anderton , C. Baus , T.P. Davis , R. Pearson , K. Mukai , J. Pollard , K. Taylor , S. Kirk , J. Hagues
Tritium self-sufficiency is one of the fundamental challenges for commercially viable deuterium–tritium nuclear fusion power stations. The combination of key high temperature radiation shielding materials that possess dense, high neutron absorption cross-section, and moderation properties, and tritium breeding materials could involve interesting design spaces for the central column challenge in spherical tokamaks. Potential tungsten alloys can be used for two functions: radiation shielding and structural material, providing a new design space window for spherical tokamak central column breeding space. In this paper, we present two novel high temperature concepts for the inboard side of the breeder blanket in a confined space, such as a spherical tokamak. A tungsten–rhenium–hafnium-carbide lithium-based design was found to offer the best TBR given a parameter optimisation based on shielding and thermal requirements. A silicon-carbide lead-lithium breeder design was also investigated. The highest TBR was found to be 0.135 in a 3D neutronics calculation with a W-24.5Re-2HfC (structural and shielding, wt%), Li (90% Li-6 enriched breeder), and tungsten pentaboride (W2B5) (shielding) option. Although this TBR is lower than unity, it will contribute to the reactor’s global TBR.
{"title":"Novel high temperature tritium blanket designs for confined spaces in spherical tokamak fusion reactors","authors":"M.D. Anderton , C. Baus , T.P. Davis , R. Pearson , K. Mukai , J. Pollard , K. Taylor , S. Kirk , J. Hagues","doi":"10.1016/j.fusengdes.2024.114732","DOIUrl":"10.1016/j.fusengdes.2024.114732","url":null,"abstract":"<div><div>Tritium self-sufficiency is one of the fundamental challenges for commercially viable deuterium–tritium nuclear fusion power stations. The combination of key high temperature radiation shielding materials that possess dense, high neutron absorption cross-section, and moderation properties, and tritium breeding materials could involve interesting design spaces for the central column challenge in spherical tokamaks. Potential tungsten alloys can be used for two functions: radiation shielding and structural material, providing a new design space window for spherical tokamak central column breeding space. In this paper, we present two novel high temperature concepts for the inboard side of the breeder blanket in a confined space, such as a spherical tokamak. A tungsten–rhenium–hafnium-carbide lithium-based design was found to offer the best TBR given a parameter optimisation based on shielding and thermal requirements. A silicon-carbide lead-lithium breeder design was also investigated. The highest TBR was found to be 0.135 in a 3D neutronics calculation with a W-24.5Re-2HfC (structural and shielding, wt%), Li (90% Li-6 enriched breeder), and tungsten pentaboride (W<sub>2</sub>B<sub>5</sub>) (shielding) option. Although this TBR is lower than unity, it will contribute to the reactor’s global TBR.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114732"},"PeriodicalIF":1.9,"publicationDate":"2024-11-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142719770","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 former ADITYA, a medium-sized tokamak with a limiter configuration was upgraded to ADITYA-U tokamak with divertor configuration. Two pairs of new divertor coils, a single pair of auxiliary divertor coils and position control coils have been introduced in ADITYA-U tokamak to achieve shaped plasma operation using the existing Toroidal field, Ohmic transformer and vertical field coils. Currently, copper-based continuous transposed conductor (CTC) has been introduced for in-situ winding of the coils. Coil insulation materials are selected to withstand high current (15 kA), high voltage (5 kV) and sustain high temperature (120 °C) during the experiment. The primary challenge was to install the coil with the bus bar using the same conductor without any joints. The design of new divertor coils mainly includes electrical and thermal considerations within the limited space available for installation. A detailed description of the installation of coils, insulation fabrication, insulation curing process and testing of the divertor coils is presented in this paper.
{"title":"Installation, thermal curing, qualification testing of divertor and position control coils in ADITYA-U tokamak","authors":"Rohit Kumar , Vaibhav Ranjan , Harshita Raj , Sharvil Patel , K. Sathyanarayana , Kaushal Patel , Kumarpal Jadeja , R.L. Tanna , J. Ghosh","doi":"10.1016/j.fusengdes.2024.114734","DOIUrl":"10.1016/j.fusengdes.2024.114734","url":null,"abstract":"<div><div>The former ADITYA, a medium-sized tokamak with a limiter configuration was upgraded to ADITYA-U tokamak with divertor configuration. Two pairs of new divertor coils, a single pair of auxiliary divertor coils and position control coils have been introduced in ADITYA-U tokamak to achieve shaped plasma operation using the existing Toroidal field, Ohmic transformer and vertical field coils. Currently, copper-based continuous transposed conductor (CTC) has been introduced for in-situ winding of the coils. Coil insulation materials are selected to withstand high current (15 kA), high voltage (5 kV) and sustain high temperature (120 °C) during the experiment. The primary challenge was to install the coil with the bus bar using the same conductor without any joints. The design of new divertor coils mainly includes electrical and thermal considerations within the limited space available for installation. A detailed description of the installation of coils, insulation fabrication, insulation curing process and testing of the divertor coils is presented in this paper.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114734"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142700896","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 : 2024-11-22DOI: 10.1016/j.fusengdes.2024.114733
D.B. Gin , B. Coriton , F. Josseaume , S. Pak , P. Ruiz , I.V. Buslakov , I.D. Kirienko , A.M. Lobachev , V.S. Modestov , I.R. Murtazin , G.E. Nemtsev , S.B. Stepanov , M.V. Ivantsivskiy , A.S. Taskaev , P.A. Seleznev , N.A. Zolotukhina , J. Kim
The study is an analysis of the strength properties of a unique diagnostic system called Upper Vertical Neutron Camera (UVNC) under the loads of various nature. The UVNC includes a DN10 water cooling system, six detector units, electrical feedthroughs, and cables. Combinations of mechanical, thermal, seismic and electromagnetic loads were considered. The implementation of multiphysical processes is carried out in ANSYS Software, and the previously obtained data on electromagnetic and seismic loads, as well as thermal fields in Normal Operation and Baking modes are used in this study as initial data for mechanical formulations. The results of the structural analysis demonstrate a sufficient margin of safety of the main structural elements of the model while maintaining operational properties. In the article described and justified totally new design of developing system under extreme neutron fluxes, high thermal loads, and combined mechanical stresses, taking into account all the unique operational factors of the ITER facility.
{"title":"Strength analysis of the Upper Vertical Neutron Camera diagnostic system of the ITER Tokamak Upper Port # 18 under electromagnetic, hydraulic, thermal, and seismic loads","authors":"D.B. Gin , B. Coriton , F. Josseaume , S. Pak , P. Ruiz , I.V. Buslakov , I.D. Kirienko , A.M. Lobachev , V.S. Modestov , I.R. Murtazin , G.E. Nemtsev , S.B. Stepanov , M.V. Ivantsivskiy , A.S. Taskaev , P.A. Seleznev , N.A. Zolotukhina , J. Kim","doi":"10.1016/j.fusengdes.2024.114733","DOIUrl":"10.1016/j.fusengdes.2024.114733","url":null,"abstract":"<div><div>The study is an analysis of the strength properties of a unique diagnostic system called Upper Vertical Neutron Camera (UVNC) under the loads of various nature. The UVNC includes a DN10 water cooling system, six detector units, electrical feedthroughs, and cables. Combinations of mechanical, thermal, seismic and electromagnetic loads were considered. The implementation of multiphysical processes is carried out in ANSYS Software, and the previously obtained data on electromagnetic and seismic loads, as well as thermal fields in Normal Operation and Baking modes are used in this study as initial data for mechanical formulations. The results of the structural analysis demonstrate a sufficient margin of safety of the main structural elements of the model while maintaining operational properties. In the article described and justified totally new design of developing system under extreme neutron fluxes, high thermal loads, and combined mechanical stresses, taking into account all the unique operational factors of the ITER facility.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114733"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701009","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 : 2024-11-20DOI: 10.1016/j.fusengdes.2024.114725
Yashan Guo , Ning Guo , Ao Zhang , Kemin Xue , Ping Li
The orientation dependence of the surface morphology for polycrystalline W after the He plasma exposure (∼ 50 eV) at ∼ 823 K was systematically investigated. Five different types of undulating morphologies were formed and the edge direction of the undulations was parallel to the 〈100〉 orientation. The relationship among crystal orientation, undulating height and undulating interval was established and the interval coefficient was introduced. The interval and the interval coefficient gradually increase until the tilting angle increases to 36 °. After severe plastic deformation (SPD), the undulating height increased from 5.7 nm for the as-received W to 11.9 nm for the SPDed W and the surface roughness of the SPDed grains increased significantly. Reducing the grains with the {100} plane and inhibiting the formation of unstable nanostructures are helpful to slow down the volume loss of W and improve its irradiation resistance.
系统研究了多晶 W 在 823 K ∼ He 等离子体暴露（∼ 50 eV）后表面形貌的取向依赖性。研究结果表明，在 823 K ∼ He 等离子体暴露（∼ 50 eV）后，W 晶体形成了五种不同的起伏形态，起伏的边缘方向与〈100〉取向平行。建立了晶体取向、起伏高度和起伏间隔之间的关系，并引入了间隔系数。起伏间隔和起伏间隔系数逐渐增大，直到倾斜角增大到 36°。经过剧烈塑性变形（SPD）后，起伏高度从原始 W 的 5.7 nm 增加到 SPD 后 W 的 11.9 nm，SPD 后晶粒的表面粗糙度也显著增加。减少带有{100}平面的晶粒和抑制不稳定纳米结构的形成有助于减缓 W 的体积损失和提高其抗辐照性能。
{"title":"Effect of severe plastic deformation on surface modification of tungsten exposed to low energy helium ion irradiation","authors":"Yashan Guo , Ning Guo , Ao Zhang , Kemin Xue , Ping Li","doi":"10.1016/j.fusengdes.2024.114725","DOIUrl":"10.1016/j.fusengdes.2024.114725","url":null,"abstract":"<div><div>The orientation dependence of the surface morphology for polycrystalline W after the He plasma exposure (∼ 50 eV) at ∼ 823 K was systematically investigated. Five different types of undulating morphologies were formed and the edge direction of the undulations was parallel to the 〈100〉 orientation. The relationship among crystal orientation, undulating height and undulating interval was established and the interval coefficient was introduced. The interval and the interval coefficient gradually increase until the tilting angle increases to 36 °. After severe plastic deformation (SPD), the undulating height increased from 5.7 nm for the as-received W to 11.9 nm for the SPDed W and the surface roughness of the SPDed grains increased significantly. Reducing the grains with the {100} plane and inhibiting the formation of unstable nanostructures are helpful to slow down the volume loss of W and improve its irradiation resistance.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114725"},"PeriodicalIF":1.9,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701015","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 : 2024-11-19DOI: 10.1016/j.fusengdes.2024.114699
Zhongtao Zhang , Zhihong Liu , Jiefeng Wu , Jianguo Ma
To ensure the stable operation of fusion reactors and synchrotron radiation facilities, the magnetic permeability of the weld joints in 316LN austenitic stainless steel (ASS) must be μr ≤ 1.03. Therefore, controlling magnetic permeability during welding is essential. This paper examines the impact of nitrogen in the shielding gas on the magnetic permeability, microstructure, and mechanical properties of 316LN welded joints. This was achieved by using argon-nitrogen mixed shielding gas to weld the test plates. The findings indicate that the primary microstructures of the welded joints consist of austenite and a small amount of ferrite. The magnetic permeabilities are measured at 1.026, 1.015, 1.010, 1.007, 1.004, and 1.005, respectively. The main factor contributing to the decline in magnetic permeability with increased nitrogen content in the shielding gas is the reduction in ferrite content. Additionally, adding nitrogen to the shielding gas improves the mechanical properties of the welded joints at 4.2 K. The ultimate tensile strengths of the welded samples using pure argon gas and argon-nitrogen mixed gas (1 % and 2 % nitrogen) were 1423.73 MPa, 1465.49 MPa, and 1546.45 MPa, respectively. The impact energy of the samples was 38 J, 139 J, and 137 J, respectively. Analysis of the fracture surfaces reveals that the argon-nitrogen mixed gas welded samples exhibit ductile fracture with dimples and micropore morphology. In contrast, the pure argon gas welded samples show the combination of ductile and brittle fracture. It is proposed that fine grain strengthening and the dispersion strengthening effect of N is the main factors in improving strength. This study provides an effective guide for the application of the welded joint of 316LN ASS at cryogenic temperature.
{"title":"The influence of nitrogen in shielding gas on the 316LN austenitic stainless steel welded joints","authors":"Zhongtao Zhang , Zhihong Liu , Jiefeng Wu , Jianguo Ma","doi":"10.1016/j.fusengdes.2024.114699","DOIUrl":"10.1016/j.fusengdes.2024.114699","url":null,"abstract":"<div><div>To ensure the stable operation of fusion reactors and synchrotron radiation facilities, the magnetic permeability of the weld joints in 316LN austenitic stainless steel (ASS) must be μ<sub>r</sub> ≤ 1.03. Therefore, controlling magnetic permeability during welding is essential. This paper examines the impact of nitrogen in the shielding gas on the magnetic permeability, microstructure, and mechanical properties of 316LN welded joints. This was achieved by using argon-nitrogen mixed shielding gas to weld the test plates. The findings indicate that the primary microstructures of the welded joints consist of austenite and a small amount of ferrite. The magnetic permeabilities are measured at 1.026, 1.015, 1.010, 1.007, 1.004, and 1.005, respectively. The main factor contributing to the decline in magnetic permeability with increased nitrogen content in the shielding gas is the reduction in ferrite content. Additionally, adding nitrogen to the shielding gas improves the mechanical properties of the welded joints at 4.2 K. The ultimate tensile strengths of the welded samples using pure argon gas and argon-nitrogen mixed gas (1 % and 2 % nitrogen) were 1423.73 MPa, 1465.49 MPa, and 1546.45 MPa, respectively. The impact energy of the samples was 38 <em>J</em>, 139 <em>J</em>, and 137 <em>J</em>, respectively. Analysis of the fracture surfaces reveals that the argon-nitrogen mixed gas welded samples exhibit ductile fracture with dimples and micropore morphology. In contrast, the pure argon gas welded samples show the combination of ductile and brittle fracture. It is proposed that fine grain strengthening and the dispersion strengthening effect of N is the main factors in improving strength. This study provides an effective guide for the application of the welded joint of 316LN ASS at cryogenic temperature.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114699"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701014","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 : 2024-11-19DOI: 10.1016/j.fusengdes.2024.114729
I. Álvarez , M. Anguiano , F. Mota , R. Hernández , F. Moro , S. Noce , Y. Qiu , J. Park , F. Arbeiter , I. Palermo , D. Sosa
The International Fusion Materials Irradiation Facility- DEMO Oriented NEutron Source (IFMIF-DONES) facility is a neutron irradiation facility specifically designed to obtain data on material irradiation for the construction of DEMO (DEMOstration) fusion power plant. The purpose of this study is to investigate and compare the effects of using different realistic models of specimen distribution during the irradiation campaign in the High Flux Test Module (HFTM) in IFMIF-DONES. Parameters such as neutron fluence rate, primary displacement damage rate and gas production have been calculated for two beam footprint sizes: the standard 20×5 cm² and the reduced 10×5 cm². The standard deuteron beam energy is 40 MeV with a current of 125 mA, but other energy values such as 25, 30 and 35 MeV have also been considered to evaluate their impact on the irradiation parameters. As the idea is to reproduce the DEMO conditions, some neutron spectra in the first wall of the DCLL, WCLL and HCPB have also been evaluated to gather reference data and compare the environments of DEMO and IFMIF-DONES. The level of packaging of the specimens impacts directly the neutron fluence rate behaviour and the different specimen distribution models give rise to different primary displacement damage rate distributions, demonstrating their versatility to meet specific needs. With respect to the comparison DEMO values, IFMIF-DONES meets the requirements of primary displacement damage rate and gas production at different beam energies. This study emphasises the essential role of sample distribution in improving the accuracy of measurements made at the IFMIF-DONES facility.
{"title":"Comparative analysis of neutronic features for various specimen payload configurations within the IFMIF-DONES HFTM","authors":"I. Álvarez , M. Anguiano , F. Mota , R. Hernández , F. Moro , S. Noce , Y. Qiu , J. Park , F. Arbeiter , I. Palermo , D. Sosa","doi":"10.1016/j.fusengdes.2024.114729","DOIUrl":"10.1016/j.fusengdes.2024.114729","url":null,"abstract":"<div><div>The International Fusion Materials Irradiation Facility- DEMO Oriented NEutron Source (IFMIF-DONES) facility is a neutron irradiation facility specifically designed to obtain data on material irradiation for the construction of DEMO (DEMOstration) fusion power plant. The purpose of this study is to investigate and compare the effects of using different realistic models of specimen distribution during the irradiation campaign in the High Flux Test Module (HFTM) in IFMIF-DONES. Parameters such as neutron fluence rate, primary displacement damage rate and gas production have been calculated for two beam footprint sizes: the standard 20×5 cm² and the reduced 10×5 cm². The standard deuteron beam energy is 40 MeV with a current of 125 mA, but other energy values such as 25, 30 and 35 MeV have also been considered to evaluate their impact on the irradiation parameters. As the idea is to reproduce the DEMO conditions, some neutron spectra in the first wall of the DCLL, WCLL and HCPB have also been evaluated to gather reference data and compare the environments of DEMO and IFMIF-DONES. The level of packaging of the specimens impacts directly the neutron fluence rate behaviour and the different specimen distribution models give rise to different primary displacement damage rate distributions, demonstrating their versatility to meet specific needs. With respect to the comparison DEMO values, IFMIF-DONES meets the requirements of primary displacement damage rate and gas production at different beam energies. This study emphasises the essential role of sample distribution in improving the accuracy of measurements made at the IFMIF-DONES facility.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114729"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701013","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-18DOI: 10.1016/j.fusengdes.2024.114723
Jing Sun , Zhaohui Yan , Xin Luo , Jiefeng Wu , Wenge Chen , Zhihong Liu
The process for Bi-2212 CICC is wind & react technology, which places very high demands on the temperature resistance of the insulation material. This study investigated the insulation structural properties of alumino-silicate fibers as insulation fibers. Initially, alumino-silicate fiber resin composites were fabricated and tested for tensile strength, interlaminar shear strength, and high voltage property. The results indicated that the alumino-silicate fiber resin composites possess relatively excellent mechanical and insulation properties. Subsequently, a mock-up, consisting of a 10 × 10 conductor array with a length of 1 meter, was constructed and subjected to a series of tests, including thermal cycling, DC, AC, and Paschen tests, adhering to ITER PF6 specifications. The results demonstrate that the mock-up meets the required criteria for both DC and AC performance. However, in the Paschen test, it exhibits a relatively low breakdown voltage, indicating areas for further optimization.
{"title":"Performance of the insulation mock-up used alumino-silicate fiber composite for Bi-2212 Cable-in-Conduit Conductor","authors":"Jing Sun , Zhaohui Yan , Xin Luo , Jiefeng Wu , Wenge Chen , Zhihong Liu","doi":"10.1016/j.fusengdes.2024.114723","DOIUrl":"10.1016/j.fusengdes.2024.114723","url":null,"abstract":"<div><div>The process for Bi-2212 CICC is wind & react technology, which places very high demands on the temperature resistance of the insulation material. This study investigated the insulation structural properties of alumino-silicate fibers as insulation fibers. Initially, alumino-silicate fiber resin composites were fabricated and tested for tensile strength, interlaminar shear strength, and high voltage property. The results indicated that the alumino-silicate fiber resin composites possess relatively excellent mechanical and insulation properties. Subsequently, a mock-up, consisting of a 10 × 10 conductor array with a length of 1 meter, was constructed and subjected to a series of tests, including thermal cycling, DC, AC, and Paschen tests, adhering to ITER PF6 specifications. The results demonstrate that the mock-up meets the required criteria for both DC and AC performance. However, in the Paschen test, it exhibits a relatively low breakdown voltage, indicating areas for further optimization.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"210 ","pages":"Article 114723"},"PeriodicalIF":1.9,"publicationDate":"2024-11-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701010","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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