2009 23rd IEEE/NPSS Symposium on Fusion Engineering最新文献_第8页

FAST Load Assembly conceptual design FAST负载总成概念设计

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226387

A. Cucchiaro, G. Brolatti, G. Calabrò, V. Cocilovo, A. Coletti, R. Coletti, P. Costa, P. Frosi, F. Crescenzi, F. Crisanti, G. Granucci, G. Maddaluno, V. Pericoli-Ridolfini, A. Pizzuto, C. Rita, G. Ramogida, S. Roccella, P. Rossi

{"title":"FAST Load Assembly conceptual design","authors":"A. Cucchiaro, G. Brolatti, G. Calabrò, V. Cocilovo, A. Coletti, R. Coletti, P. Costa, P. Frosi, F. Crescenzi, F. Crisanti, G. Granucci, G. Maddaluno, V. Pericoli-Ridolfini, A. Pizzuto, C. Rita, G. Ramogida, S. Roccella, P. Rossi","doi":"10.1109/FUSION.2009.5226387","DOIUrl":"https://doi.org/10.1109/FUSION.2009.5226387","url":null,"abstract":"FAST (Fusion Advanced Studies Torus) is a proposal for a Satellite Facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploiting innovative DEMO technology. FAST is a compact (Ro = 1.82 m, a = 0.64 m, triangularity δ = 0.4) machine able to investigate non linear dynamics effects of alpha particle behaviours in burning plasmas [1][ [2][5]. The project is based on a dominant 30 MW of ICRH, 6 MW of LH and 4 MW of ECRH. FAST operates at a wide range [3][4] of parameters e.g., in high performance H-Mode (BT up to 8.5 T; IP up to 8 MA) as well as in Advanced Tokamak operation (IP=3 MA), and full non inductive current scenario (IP=2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets [6]. That allows for a pulse duration up to 170 s. Tungsten (W) or Liquid Lithium (L-Li) have been chosen as the divertor plates material, and Argon or Neon as the injected impurities to mitigate the thermal loads.","PeriodicalId":236460,"journal":{"name":"2009 23rd IEEE/NPSS Symposium on Fusion Engineering","volume":"114 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2009-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124062885","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Conceptual engineering method for attenuating He ion interactions on first wall components in the Fusion Test Facility (FTF) employing a low-pressure noble gas 利用低压惰性气体衰减聚变试验装置(FTF)第一壁组件上He离子相互作用的概念工程方法

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226438

C. Gentile, W. Blanchard, T. Kozub, C. Priniski, I. Zatz, S. Obenschain

{"title":"Conceptual engineering method for attenuating He ion interactions on first wall components in the Fusion Test Facility (FTF) employing a low-pressure noble gas","authors":"C. Gentile, W. Blanchard, T. Kozub, C. Priniski, I. Zatz, S. Obenschain","doi":"10.1109/FUSION.2009.5226438","DOIUrl":"https://doi.org/10.1109/FUSION.2009.5226438","url":null,"abstract":"It has been shown that post detonation energetic helium ions can drastically reduce the useful life of the (dry) first wall of an IFE reactor due to the accumulation of implanted helium. For the purpose of attenuating energetic helium ions from interacting with first wall components in the Fusion Test Facility (FTF) target chamber, several concepts have been advanced. These include magnetic intervention (MI), deployment of a dynamically moving first wall, use of a sacrificial shroud, designing the target chamber large enough to mitigate the damage caused by He ions on the target chamber wall, and the use of a low pressure noble gas resident in the target chamber during pulse power operations. It is proposed that employing a low-pressure (∼ 1 torr equivalent) noble gas in the target chamber will thermalize energetic helium ions prior to interaction with the wall. The principle benefit of this concept is the simplicity of the design and the utilization of (modified) existing technologies for pumping and processing the noble ambient gas. Although the gas load in the system would be increased over other proposed methods, the use of a “gas shield” may provide a cost effective method of greatly extending the first wall of the target chamber. An engineering study has been initiated to investigate conceptual engineering methods for implementing a viable gas shield strategy in the FTF.","PeriodicalId":236460,"journal":{"name":"2009 23rd IEEE/NPSS Symposium on Fusion Engineering","volume":"93 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2009-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127402888","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

ITER central solenoid alternate pre-compression structure - Concept and analysis ITER中央螺线管交替预压缩结构。概念与分析

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226489

P. Litherland, K. Freudenberg, R. Myatt

{"title":"ITER central solenoid alternate pre-compression structure - Concept and analysis","authors":"P. Litherland, K. Freudenberg, R. Myatt","doi":"10.1109/FUSION.2009.5226489","DOIUrl":"https://doi.org/10.1109/FUSION.2009.5226489","url":null,"abstract":"The ITER Central Solenoid (CS) is comprised of six independent coils held together by a pre-compression support structure. This structure must provide enough preload to maintain sufficient coil-to-coil contact throughout the current pulse. End of burn (EOB) represents one extreme time-point when the currents in the CS3 coils oppose those of CS1 & CS2. The present baseline design uses nine tie plate pairs located on the coil ID and OD to achieve this pre-compression. While this is an efficient structural approach, the outer tie plates limit access to the coil OD and result in very tight tolerances over the 12.75 m tall toroidal field (TF) coil bore. An alternative pre-compression structure has been developed by the US ITER Project Office, which addresses both of these issues using only internal tension members. This allows clear access to the outer CS surface for plumbing and current feeders, and more reasonable OD tolerances. In addition, the Tension Rod pre-compression structure utilizes existing technology (such as hydraulic tensioner or Superbolt®), which streamlines initial preloading and subsequent adjustments. Structural and transient electromagnetic FEA of the tension rod concept demonstrates that the stresses, displacements, and preliminary Eddy current heat loads are within allowable limits. Fabrication estimates demonstrate no additional cost of the tension rods compared to the tie plates.","PeriodicalId":236460,"journal":{"name":"2009 23rd IEEE/NPSS Symposium on Fusion Engineering","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2009-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127945305","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

Completing the viability demonstration of direct-drive inertial fusion energy target engagement 完成直接驱动惯性聚变能目标交战可行性论证

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226375

L. Carlson, M. Tillack, J. Stromsoe, Neil Alexander, G. W. Flint, D. Goodin, R. Petzoldt

引用次数: 1

A Pfd-based 1D model for dynamic and transient tritium transfers between ITER HCLL TBM auxiliary systems 基于pfd的氚在ITER HCLL TBM辅助系统间动态和瞬态转移的一维模型

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226409

C. Moreno, L. Sedano

引用次数: 9

Combustion of hydrogen from the ITER Water Detritiation System ITER水分解系统中氢的燃烧

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226485

W. Wurster, I. Cristescu, F. Eichelhardt, G. Ana, P. Pfeifer

{"title":"Combustion of hydrogen from the ITER Water Detritiation System","authors":"W. Wurster, I. Cristescu, F. Eichelhardt, G. Ana, P. Pfeifer","doi":"10.1109/FUSION.2009.5226485","DOIUrl":"https://doi.org/10.1109/FUSION.2009.5226485","url":null,"abstract":"In nuclear fusion technology, as it is applied in the fusion reactor ITER, tritium is used as fuel to promote deuterium - tritium fusion reactions. As a result of tritium handling, considerable amounts of tritiated water are produced in the different subsystems of the reactor. Therefore, the ITER Tritium Plant will be equipped with a Water Detritiation System for the recovery of tritium from water. Tritium will be recovered as fuel and its release to the environment has to be minimized. During the operation of the ITER Water Detritiation System up to 150 m3/h hydrogen will be produced. Tritium is removed from the hydrogen stream down to an environmentally compatible concentration. The hydrogen can not be further used; therefore it is intended to dispose of it by a simple, safe and proven process. This paper presents a study of different process options: a burner with combustion chamber, catalytic combustion in a conventional catalyst bed, catalytic combustion in a passive autocatalytic recombiner and catalytic combustion in a microchannel reactor. The options are compared under the aspects of safety, investment costs and costs of operation and maintenance. A proposal for ITER will be presented.","PeriodicalId":236460,"journal":{"name":"2009 23rd IEEE/NPSS Symposium on Fusion Engineering","volume":"344 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2009-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123102050","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 2

Updated design of the ITER magnet system gravity supports ITER磁体系统重力支撑的更新设计

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226496

R. Gallix, Y. Fu, C. Jong, P. Lee, B. Hou, G. Jian

引用次数: 0

Safety related R&D for the ITER baseline design ITER基线设计的安全相关研发

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226518

S. Reyes, N. Taylor, P. Cortes, S. Ciattaglia, S. Rosanvallon, A. Perevezentsev, M. Iseli, D. Baker, J. Elbez-Uzan, L. Topilski, W. Gulden, P. Sharpe, T. Hayashi

引用次数: 0

Real-time reconstruction of the magnetic flux in ftu using multipolar current moments 利用多极电流矩实时重建ftu磁通量

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226480

Y. Sadeghi, L. Zaccarian, C. D'Epifanio, G. Ramogida, L. Boncagni, E. Vitale, F. Crisanti

引用次数: 0

Superconducting magnet power supply system for the KSTAR 1st plasma experiment and engineering KSTAR一号等离子体实验与工程超导磁体供电系统

2009 23rd IEEE/NPSS Symposium on Fusion Engineering Pub Date : 2009-06-01 DOI: 10.1109/FUSION.2009.5226386

J.H. Choi, C.H. Kim, D.K. Lee, K.H. Kim, S. Hahn, Y. Chu, J. Kong, S. L. Hong, J.K. Jin, I. Hwang, Y.S. Kim, J. Park, M. Kwon, H. Ahn, G. Jang, M. Yun, D. Seong, Y. Kim, Y.W. Lee, H.S. Shin

{"title":"Superconducting magnet power supply system for the KSTAR 1st plasma experiment and engineering","authors":"J.H. Choi, C.H. Kim, D.K. Lee, K.H. Kim, S. Hahn, Y. Chu, J. Kong, S. L. Hong, J.K. Jin, I. Hwang, Y.S. Kim, J. Park, M. Kwon, H. Ahn, G. Jang, M. Yun, D. Seong, Y. Kim, Y.W. Lee, H.S. Shin","doi":"10.1109/FUSION.2009.5226386","DOIUrl":"https://doi.org/10.1109/FUSION.2009.5226386","url":null,"abstract":"The Korea Superconducting Tokamak Advanced Research (KSTAR) device is an advanced superconducting tokamak to establish scientific and technological bases for an attractive fusion reactor. This device requires 3.5 Tesla of toroidal field (TF) for plasma confinement, and requires a strong poloidal flux swing to generate an inductive voltage to produce and sustain the tokamak plasma. KSTAR was originally designed to have 16 serially connected TF magnets for which the nominal current rating is 35.2 kA. KSTAR also has 7 pairs of poloidal field (PF) coils that are driven to 1 MA/sec for generation of the tokamak plasma according to the operation scenarios. The KSTAR Magnet Power Supply (MPS) was dedicated to the superconducting (SC) coil commissioning and 1st plasma experiment as a part of the system commissioning. This paper will describe key features of KSTAR MPS for the 1st plasma experiment, and will also report the engineering and commissioning results of the magnet power supplies.","PeriodicalId":236460,"journal":{"name":"2009 23rd IEEE/NPSS Symposium on Fusion Engineering","volume":"45 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2009-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124090758","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 2