Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231)

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The DIII-D Neutral Beam Supervisory Control and Data Acquisition workstation upgrade
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 36-39
K.H. Doan, J.L. Busath, D.H. Kellman
The DIII-D Neutral Beam Supervisory Control and Data Acquisition (NB SCADA) system is responsible for data and status communication between remote system devices. Some years ago, it was operated and controlled on a 486 PC with Microsoft Windows 3.1. A 16-bit software package called FIXDMACS was used to interface and communicate with Siemens programmable logic controllers (PLCs). Due to the ever-ch...... hiện toàn bộ
#SCADA systems #Workstations #Niobium #Programmable control #Hardware #Software performance #Control systems #Communication system control #Software packages #Computer industry
Compact probe design for power monitoring from the narrow side of the reduced height waveguide
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 268-271
C. Kung, S. Bernabei, J. Gumbas, N. Greenough, E. Fredd
Due to confined space in a stack of reduced height waveguides, power detection of the incident and reflected wave in the reduced height waveguide is extremely difficult. A new compact probe to monitor the incident and reflected wave from the narrow side of the reduced height waveguide has been developed. This compact probe consists of 2 current loops, a directional coupler, and a small trimmer cap...... hiện toàn bộ
#Probes #Monitoring #Directional couplers #Impedance #Power transmission lines #Plasma confinement #Capacitors #Testing #Tokamaks #Power cables
Selection of plasma-facing materials in next-step fusion devices
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 311-320
G. Federici, V. Barabash, G. Janeschitz, R. Tivey, A. Loarte, J. Roth, C.H. Skinner, W.R. Wampler
Designing a robust interface between a thermonuclear plasma and the solid material environment remains a major challenge for next-step fusion devices and future power fusion reactors. Challenging trade-offs in material choice for plasma-facing components were identified in the internationally co-ordinated R&D program supporting the Engineering Design Activities of the ITER project. This paper surv...... hiện toàn bộ
#Plasma devices #Plasma materials processing #Fusion reactor design #Plasma applications #Robustness #Solids #Fusion reactors #Research and development #Design engineering #Power engineering and energy
Engineering and enabling technology development for heavy ion fusion drivers
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 484-486
W. Waldron, B. Logan, L. Ahle, G. Sabbi
The Heavy Ion Fusion Virtual National Laboratory is a collaboration among LBNL, LLNL, and PPPL. The engineering and technology development activities arc closely aligned with the major experimental areas, which include injectors, beam transport, and final focus. High current density ion sources to produce a more compact multiple beam injector are a major focus of the current activities. There are ...... hiện toàn bộ
#Collaboration #Manufacturing #Pulse compression methods #Magnetic materials #Superconducting magnets #Laboratories #Current density #Ion sources #Insulation #Costs
Post-test calculations with ISAS-ITER system for ICE experiments
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 48-51
M.T. Porfiri, P. Meloni
In the frame of the Safety and Environment tasks of the European Technology Program for ITER project one of the main issue is the validation of the computer codes and models used as reference for ITER safety analysis to obtain acceptance by licensing authorities. In the context of the fusion field facilities the data useful for validation are very limited because only few experimental machines are...... hiện toàn bộ
#Ice #Safety #Testing #Licenses #Coolants #Computational modeling #Computer simulation #Containers #Plasma simulation #Analytical models
Structure analysis and experiment research of the welded bellows for the ports of the HT-7U vacuum vessel
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 376-379
Yuntao Song, Damao Yao, Songtao Wu, Jie Yu, Peide Weng, Yihua Liu, Jionghua Wang, Xuetao Wu
Vacuum vessel of the HT-7U is a fully welded toroidal structure with noncircular cross-section nested in the bore of the TF coils. According to the requirement of the physics design, sixteen horizontal ports on outboard mid-plane and thirty-two vertical ports on the top and bottom are designed for diagnostics, plasma heating, current driving, vacuum pumping and gas puffing. Bellows on these ports ...... hiện toàn bộ
#Welding #Bellows #Fatigue #Testing #Finite element methods #Performance analysis #Prototypes #Boring #Coils #Physics
A toroidal liquid lithium limiter for CDX-U
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 341-344
R. Majeski, G. Antar, M. Boaz, D. Buchenauer, L. Cadwallader, R. Causey, R.W. Conn, R. Doerner, P. Efthimion, M. Finkenthal, D. Hoffman, B. Jones, R. Kaita, H. Kugel, S. Luckhardt, R. Maingi, M. Maiorano, T. Munsat, S. Raftopoulos, T. Rognlein, J. Spaleta, V. Soukhanovskii, D. Stutman, G. Taylor, J. Timberlake, M. Ulrickson, D. Whyte
Attention has focused recently on flowing liquid lithium as a first wall for a reactor because of its potentially attractive physics and engineering features. In order to test the suitability of liquid lithium as a plasma facing component, the Current Drive eXperiment - Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has recently installed a fully toroidal liquid lithium limiter. CDX-U ...... hiện toàn bộ
#Lithium #Physics #Plasma diagnostics #Inductors #Testing #Laboratories #Steel #Surface discharges #Heating #Rails
High heat flux interactions and tritium removal from plasma facing components by a scanning laser
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 473-476
C.H. Skinner, C.A. Gentile, A. Hassanein
A new technique for studying high heat flux interactions with plasma facing components is presented. The beam from a continuous wave 300 W neodymium laser was focussed to 80 W/mm/sup 2/ and scanned at high speed over the surface of carbon tiles. These tiles were previously used in the TFTR inner limiter and have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma op...... hiện toàn bộ
#Plasma temperature #Tiles #Plasma waves #Manufacturing #Laser beams #Neodymium #Surface waves #Surface emitting lasers #Amorphous materials #Plasma materials processing
Thermal response and thermal-hydraulic analysis of PFC baking for SST-1 tokamak
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 469-472
P. Chaudhuri, D.C. Reddy, S. Khirwadkar, N.R. Prakash, P. Santra, Y.C. Saxena
Steady state Superconducting Tokamak (SST-1) to address some of the physics and technology issues related to steady state tokamak operation. The plasma facing components (PFC) of SST-1, placed inside the vacuum vessel (VV) of the tokamak, are designed to be compatible for steady state operation. The main consideration in the design of the PFC is the steady state heat removal of up to 1 MW/m/sup 2/...... hiện toàn bộ
#Tokamaks #Steady-state #Plasma temperature #High temperature superconductors #Superconducting coils #Nitrogen #Thermal loading #Impurities #Heat transfer #Physics
Những tiến bộ gần đây từ chương trình DIII-D Dịch bởi AI
Proceedings of the 19th IEEE/IPSS Symposium on Fusion Engineering. 19th SOFE (Cat. No.02CH37231) - - Trang 442-447
A.G. Kellman
Tiến bộ đáng kể đã được thực hiện trong một số lĩnh vực khoa học và kỹ thuật chính, rất quan trọng cho hoạt động tokamak tiên tiến trên tokamak DIII-D. Việc cải thiện sửa lỗi trường động liên kết với sự quay plasma đã dẫn đến một quá trình phóng điện ổn định thụ động với bức tường ở mức gấp đôi giới hạn beta không có bức tường. Ổn định phản hồi chủ động của chế độ tường điện trở (RWM) đã được cải ...... hiện toàn bộ
#Đo plasma #Tokamak #Phản hồi #Cuộn dây #Electron #Hệ thống điều khiển #Sửa lỗi #Cảm biến từ tính #Cyclotron #Mô hình dự đoán
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