Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027635
J. Zaks, M. DeMaria, B. LaBombard, R. Granetz, E. Fitzgerald, H. Savelli, P. Stahle
This paper covers in detail the design and manufacturing of the new inner divertor for the Alcator C-Mod tokamak. We focus on the complexity of modeling the inner divertor components and the challenges of its fabrication. Analysis of the inner divertor stress and thermal expansion is presented in another paper in this conference. The model and mock-up of the inner divertor are shown.
{"title":"Inner divertor detailed design and manufacturing","authors":"J. Zaks, M. DeMaria, B. LaBombard, R. Granetz, E. Fitzgerald, H. Savelli, P. Stahle","doi":"10.1109/FUSION.2002.1027635","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027635","url":null,"abstract":"This paper covers in detail the design and manufacturing of the new inner divertor for the Alcator C-Mod tokamak. We focus on the complexity of modeling the inner divertor components and the challenges of its fabrication. Analysis of the inner divertor stress and thermal expansion is presented in another paper in this conference. The model and mock-up of the inner divertor are shown.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90212427","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027642
S. Cox, A. Bickley, T. Jones, J. Milnes
Physics modelling and engineering analysis have been carried out to determine the operating limits of the upgraded JET neutral beam injector from duct re-ionisation and beam shine-through. The JET neutral beam duct is only 23cm wide and 90cm tall at its throat and yet it presently has to transmit more than 11MW of D/sup 0/ beam particles, resulting in power densities in excess of 200MW/m/sup 2/. Even at this power level, the copper duct liner can be the limiting component with respect to the pulse length of the Octant 4 injector, depending on plasma current and power. The upgrade to the Octant 8 injector in 2002 will increase the power to /spl sim/15MW of D/sup 0/ at 130kV, so it is necessary to determine the new limits. It is shown that at full power, the duct will become the major limiting component with respect to pulse length for this injector. The shine-through power density and integrated energy for various in-vessel components have also been evaluated for the upgraded injector. Thermo-mechanical finite element stress calculations on elements of the ICRH antenna show that the injector can be operated at full power without further restrictions being imposed on the plasma characteristics, e.g. density and shape. For the CFC inner wall guard limiter tiles and their internal reinforcement, however, there is a bulk temperature limit and an enhancement to the existing real-time protection system is proposed.
{"title":"Operating limits of the upgraded JET neutral beam injector from duct re-ionisation and beam shine-through","authors":"S. Cox, A. Bickley, T. Jones, J. Milnes","doi":"10.1109/FUSION.2002.1027642","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027642","url":null,"abstract":"Physics modelling and engineering analysis have been carried out to determine the operating limits of the upgraded JET neutral beam injector from duct re-ionisation and beam shine-through. The JET neutral beam duct is only 23cm wide and 90cm tall at its throat and yet it presently has to transmit more than 11MW of D/sup 0/ beam particles, resulting in power densities in excess of 200MW/m/sup 2/. Even at this power level, the copper duct liner can be the limiting component with respect to the pulse length of the Octant 4 injector, depending on plasma current and power. The upgrade to the Octant 8 injector in 2002 will increase the power to /spl sim/15MW of D/sup 0/ at 130kV, so it is necessary to determine the new limits. It is shown that at full power, the duct will become the major limiting component with respect to pulse length for this injector. The shine-through power density and integrated energy for various in-vessel components have also been evaluated for the upgraded injector. Thermo-mechanical finite element stress calculations on elements of the ICRH antenna show that the injector can be operated at full power without further restrictions being imposed on the plasma characteristics, e.g. density and shape. For the CFC inner wall guard limiter tiles and their internal reinforcement, however, there is a bulk temperature limit and an enhancement to the existing real-time protection system is proposed.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74326920","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027669
Y. Kawarnata, I. Yonekawa, K. Kurihara
A new digital integrator with high input voltage and a long-time low drift speed has been developed in JT-60. This integrator uses an integrating method with a pair of a voltage-to-frequency converter (VFC) and an up and down counter (UDC) (VFC-UDC unit). However, this method has a saturation of the VFC for large input, leading to integration error. To improve it, the new digital integrator is composed of three sets of VFC-UDC units in parallel and a digital signal processor (DSP). Three VFC-UDC units with different input ranges integrate an identical input signal respectively, and the DSP selects a suitable integrated signal among three integrated outputs at a sampling frequency of 10 kHz and makes a chain of integrated signals. The performance of the integrator has been tested using a disruptive discharge in JT-60. A good integration result has been obtained though large signal is input. Also, fundamental characteristics of the VFC, linearity, thermal drift, dead band, which are causes of integration errors, are described.
{"title":"Development of an intelligent digital integrator for long-pulse operation in a tokamak","authors":"Y. Kawarnata, I. Yonekawa, K. Kurihara","doi":"10.1109/FUSION.2002.1027669","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027669","url":null,"abstract":"A new digital integrator with high input voltage and a long-time low drift speed has been developed in JT-60. This integrator uses an integrating method with a pair of a voltage-to-frequency converter (VFC) and an up and down counter (UDC) (VFC-UDC unit). However, this method has a saturation of the VFC for large input, leading to integration error. To improve it, the new digital integrator is composed of three sets of VFC-UDC units in parallel and a digital signal processor (DSP). Three VFC-UDC units with different input ranges integrate an identical input signal respectively, and the DSP selects a suitable integrated signal among three integrated outputs at a sampling frequency of 10 kHz and makes a chain of integrated signals. The performance of the integrator has been tested using a disruptive discharge in JT-60. A good integration result has been obtained though large signal is input. Also, fundamental characteristics of the VFC, linearity, thermal drift, dead band, which are causes of integration errors, are described.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72572119","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027663
T. Mito, A. Nishimura, S. Yamada, S. Imagawa, K. Takahata, N. Yanagi, R. Maekawa, H. Chikaraishi, H. Tamura, A. Iwamoto, S. Hamaguchi, Y. Hishinuma, T. Satow, O. Motojima
The Large Helical Device (LHID) of National Institute for Fusion Science (NIFS) is a heliotron-type experimental fusion device which has the capability of confining current-less and steady-state plasma. The primary feature on the engineering aspect of LHD is using superconducting coils for magnetic confinement: two pool boiling helical coils (H1, H2) and three pairs of forced-flow poloidal coils (IV, IS, OV) wound with cable-in-conduit conductors (CICC). The maximum magnetic field at plasma center is 3 T in the Phase I experiment and 4 T in Phase II, while its stored energy becomes 0.9 GJ and 1.6 GJ, respectively. These coils are connected to the power supplies by superconducting bus-lines with their nominal current of 31.3 kA. The construction of LHD started in 1991 and was completed by the end of 1997. During this period, extensive research and development were conducted to complete a large-scale superconducting system. The plasma experiment started on March 31, 1998 and four plasma experimental campaigns have been performed successfully in three years. The fifth cycle operation started in August 2001. The knowledge which has been acquired during the design, development, and operation of superconducting system for LHD, is summarized.
美国国家聚变科学研究所(NIFS)的大型螺旋装置(LHID)是一种具有约束无电流稳态等离子体的氦氖型实验聚变装置。LHD工程方面的主要特点是使用超导线圈进行磁约束:两个池沸腾螺旋线圈(H1, H2)和三对强制流动极向线圈(IV, is, OV)，缠绕着电缆导管导体(CICC)。第一期实验等离子体中心最大磁场为3 T，第二期实验为4 T，其存储能量分别为0.9 GJ和1.6 GJ。这些线圈通过超导母线连接到电源，其标称电流为31.3 kA。LHD的建设于1991年开始，并于1997年底完成。在此期间，进行了广泛的研究和开发，以完成大规模的超导系统。等离子体实验于1998年3月31日开始，三年内成功进行了四次等离子体实验。第五个周期运作于2001年8月展开。总结了LHD超导系统在设计、开发和运行过程中所获得的知识。
{"title":"Design, development and operation of superconducting system for LHD","authors":"T. Mito, A. Nishimura, S. Yamada, S. Imagawa, K. Takahata, N. Yanagi, R. Maekawa, H. Chikaraishi, H. Tamura, A. Iwamoto, S. Hamaguchi, Y. Hishinuma, T. Satow, O. Motojima","doi":"10.1109/FUSION.2002.1027663","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027663","url":null,"abstract":"The Large Helical Device (LHID) of National Institute for Fusion Science (NIFS) is a heliotron-type experimental fusion device which has the capability of confining current-less and steady-state plasma. The primary feature on the engineering aspect of LHD is using superconducting coils for magnetic confinement: two pool boiling helical coils (H1, H2) and three pairs of forced-flow poloidal coils (IV, IS, OV) wound with cable-in-conduit conductors (CICC). The maximum magnetic field at plasma center is 3 T in the Phase I experiment and 4 T in Phase II, while its stored energy becomes 0.9 GJ and 1.6 GJ, respectively. These coils are connected to the power supplies by superconducting bus-lines with their nominal current of 31.3 kA. The construction of LHD started in 1991 and was completed by the end of 1997. During this period, extensive research and development were conducted to complete a large-scale superconducting system. The plasma experiment started on March 31, 1998 and four plasma experimental campaigns have been performed successfully in three years. The fifth cycle operation started in August 2001. The knowledge which has been acquired during the design, development, and operation of superconducting system for LHD, is summarized.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84205479","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027674
K. Young
The physics program of any next step tokamak such as FIRE sets demands for plasma measurement which are at least as comprehensive as on present tokamaks, with the additional capabilities needed for control of the plasma and for understanding the effects of the alpha-particles. The diagnostic instrumentation must be able to provide the fine spatial and temporal resolution required for the advanced tokamak plasma scenarios. It must also be able to overcome the effects of neutron- and gamma-induced electrical noise in ceramic components or detectors, and fluorescence and absorption in optical components. There are practical engineering issues of minimizing radiation streaming while providing essential diagnostic access to the plasma. Many diagnostics will require components at or close to the first wall, e.g. ceramics and MI cable for magnetic diagnostics and mirrors for optical diagnostics; these components must be mounted to operate, and survive, in fluxes which require special material selection. A better set of diagnostics of alpha-particles than that available for TFTR is essential; it must be qualified well before moving into D-T experiments. A start has been made to assessing the potential implementation of key diagnostics for the FIRE device. The present status is described.
{"title":"Challenges for plasma diagnostics in a next step device (FIRE)","authors":"K. Young","doi":"10.1109/FUSION.2002.1027674","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027674","url":null,"abstract":"The physics program of any next step tokamak such as FIRE sets demands for plasma measurement which are at least as comprehensive as on present tokamaks, with the additional capabilities needed for control of the plasma and for understanding the effects of the alpha-particles. The diagnostic instrumentation must be able to provide the fine spatial and temporal resolution required for the advanced tokamak plasma scenarios. It must also be able to overcome the effects of neutron- and gamma-induced electrical noise in ceramic components or detectors, and fluorescence and absorption in optical components. There are practical engineering issues of minimizing radiation streaming while providing essential diagnostic access to the plasma. Many diagnostics will require components at or close to the first wall, e.g. ceramics and MI cable for magnetic diagnostics and mirrors for optical diagnostics; these components must be mounted to operate, and survive, in fluxes which require special material selection. A better set of diagnostics of alpha-particles than that available for TFTR is essential; it must be qualified well before moving into D-T experiments. A start has been made to assessing the potential implementation of key diagnostics for the FIRE device. The present status is described.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82074587","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027682
M. Kalish, M. Cropper, C. Neumeyer, R. Parsells, L. Dudek, A. Klink, L. Morris
NSTX requires that its internal plasma facing components (PFCs) reach 350/spl deg/C during "bakeout" conditioning of the vacuum vessel. This paper describes a helium system designed to meet this requirement as well as provide cooling during plasma operations. The NSTX vacuum vessel's PFCs were designed to be heated or cooled by flowing a fluid medium through tubing attached to the PFC's copper backing plates. The heating/cooling system must move enough fluid at a sufficient rate with a high enough heat capacity through these restrictive paths. After the evaluation of several approaches including the use of heat transfer oils and steam, a compressed helium system was determined to be the optimal choice. The helium system utilizes a blower operating inside of a pressure vessel. This arrangement allows the base pressure to be raised to 20 atmospheres. With the system pressure elevated, the helium blower need only provide the motive force for overcoming 28 psi of friction losses and is not encumbered with compressing the gas. At 20 atmospheres the density of the helium is high enough to provide the heat capacity necessary to meet the NSTX requirements of 66 kW for heating and 82 kW for cooling. The paper will detail the unique design problems associated with a high pressure high temperature helium system as well as review the overall design, and modes of operation.
{"title":"Design of the NSTX heating and cooling system","authors":"M. Kalish, M. Cropper, C. Neumeyer, R. Parsells, L. Dudek, A. Klink, L. Morris","doi":"10.1109/FUSION.2002.1027682","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027682","url":null,"abstract":"NSTX requires that its internal plasma facing components (PFCs) reach 350/spl deg/C during \"bakeout\" conditioning of the vacuum vessel. This paper describes a helium system designed to meet this requirement as well as provide cooling during plasma operations. The NSTX vacuum vessel's PFCs were designed to be heated or cooled by flowing a fluid medium through tubing attached to the PFC's copper backing plates. The heating/cooling system must move enough fluid at a sufficient rate with a high enough heat capacity through these restrictive paths. After the evaluation of several approaches including the use of heat transfer oils and steam, a compressed helium system was determined to be the optimal choice. The helium system utilizes a blower operating inside of a pressure vessel. This arrangement allows the base pressure to be raised to 20 atmospheres. With the system pressure elevated, the helium blower need only provide the motive force for overcoming 28 psi of friction losses and is not encumbered with compressing the gas. At 20 atmospheres the density of the helium is high enough to provide the heat capacity necessary to meet the NSTX requirements of 66 kW for heating and 82 kW for cooling. The paper will detail the unique design problems associated with a high pressure high temperature helium system as well as review the overall design, and modes of operation.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84631461","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027678
C. Talarico, A. Lo Bue, L. Semeraro, F. Hurd, C. Cassani
The Toroidal Field Model Coil (TFMC) is a 36 tons "race-track" shaped coil about 4 m high and 3 m wide scaled with respect to the full-size ITER TF coils and including the key technical features and manufacturing approaches foreseen for the actual ITER TF coils. The objective of the TFMC Project is to develop and demonstrate the superconducting magnet technology to a level that will allow the ITER TF coils to be built with confidence. The TFMC has been installed in the TOSKA facility (Forschunszentrum Karlsruhe, Germany) after assembling the TFMC into a 27 tons stainless steel Intercoil Structure which interfaces the TFMC to the TOSKA facility. Due to the complexity of the assembly operations, and in order to avoid any trial-and-error assembly process with the associated loss of time and waste of resources, computer simulations have been extensively used. Additionally, starting from 3D CATIA models, a laser tracking technique has been utilized to retrieve as built geometry data of each sub assembly. The raw data have been analyzed and combined to verify the assembly procedure and to identify corrective actions before the real installation. The whole survey has been accomplished by ENEA during three different survey campaigns carried out at Alstom (Belfort, France) and at Forschungszentrum Karlsruhe (FZK, Karlsruhe, Germany) sites. The implementation of the data resulted in the TFMC being installed with no loss of time due to modifications to components previously measured and analyzed by the method reported here. This paper illustrates the geometry survey, the method and the instrumentation adopted and the results obtained. Moreover, it gives guidelines to the designer to be taken into account in the assembly planning of large and heavy components.
{"title":"Toroidal Field Model Coil geometry survey","authors":"C. Talarico, A. Lo Bue, L. Semeraro, F. Hurd, C. Cassani","doi":"10.1109/FUSION.2002.1027678","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027678","url":null,"abstract":"The Toroidal Field Model Coil (TFMC) is a 36 tons \"race-track\" shaped coil about 4 m high and 3 m wide scaled with respect to the full-size ITER TF coils and including the key technical features and manufacturing approaches foreseen for the actual ITER TF coils. The objective of the TFMC Project is to develop and demonstrate the superconducting magnet technology to a level that will allow the ITER TF coils to be built with confidence. The TFMC has been installed in the TOSKA facility (Forschunszentrum Karlsruhe, Germany) after assembling the TFMC into a 27 tons stainless steel Intercoil Structure which interfaces the TFMC to the TOSKA facility. Due to the complexity of the assembly operations, and in order to avoid any trial-and-error assembly process with the associated loss of time and waste of resources, computer simulations have been extensively used. Additionally, starting from 3D CATIA models, a laser tracking technique has been utilized to retrieve as built geometry data of each sub assembly. The raw data have been analyzed and combined to verify the assembly procedure and to identify corrective actions before the real installation. The whole survey has been accomplished by ENEA during three different survey campaigns carried out at Alstom (Belfort, France) and at Forschungszentrum Karlsruhe (FZK, Karlsruhe, Germany) sites. The implementation of the data resulted in the TFMC being installed with no loss of time due to modifications to components previously measured and analyzed by the method reported here. This paper illustrates the geometry survey, the method and the instrumentation adopted and the results obtained. Moreover, it gives guidelines to the designer to be taken into account in the assembly planning of large and heavy components.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84016913","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027640
H. Ahn, Y.W. Lee, Y. Kim, J. Bak, G.S. Lee
The objectives of this analysis are to investigate the structural integrity of the KSTAR (Korea Superconducting Tokamak Advanced Research) device for the design earthquake and to enhance its aseismic performance. Three-dimensional finite element beam-shell models of the vacuum vessel, cryostat, and pumping ducts were developed and spectrum analyses based on the floor response spectra for the KSTAR building were carried out separately for each finite element model. Natural frequencies, seismic displacements and stress intensities were determined and effective seismic accelerations for the static analysis were also calculated by using the reaction forces at the fixed points. From a comparative study on the two kinds of seismic analyses (spectrum and static), equivalent static earthquake loads were verified for the detailed structural analysis. The results reveal that effective seismic accelerations in the horizontal direction are three times greater than the peak ground acceleration (PGA) and those in the vertical direction are less than two times of PGA. It can also be shown that the maximum stress intensities are less than 20% of the allowable stress limit specified in the ASME Boiler and Pressure Vessel Code.
{"title":"Seismic analysis of the KSTAR tokamak","authors":"H. Ahn, Y.W. Lee, Y. Kim, J. Bak, G.S. Lee","doi":"10.1109/FUSION.2002.1027640","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027640","url":null,"abstract":"The objectives of this analysis are to investigate the structural integrity of the KSTAR (Korea Superconducting Tokamak Advanced Research) device for the design earthquake and to enhance its aseismic performance. Three-dimensional finite element beam-shell models of the vacuum vessel, cryostat, and pumping ducts were developed and spectrum analyses based on the floor response spectra for the KSTAR building were carried out separately for each finite element model. Natural frequencies, seismic displacements and stress intensities were determined and effective seismic accelerations for the static analysis were also calculated by using the reaction forces at the fixed points. From a comparative study on the two kinds of seismic analyses (spectrum and static), equivalent static earthquake loads were verified for the detailed structural analysis. The results reveal that effective seismic accelerations in the horizontal direction are three times greater than the peak ground acceleration (PGA) and those in the vertical direction are less than two times of PGA. It can also be shown that the maximum stress intensities are less than 20% of the allowable stress limit specified in the ASME Boiler and Pressure Vessel Code.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87475725","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027695
Weng Peide
The HT-7U is a superconducting tokamak, which is constructing in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). The mission of the HT-7U project is to develop scientific and engineering issues on the steady state operation of advanced tokamak. The engineering design of the device is optimized. The R&D program is going on. Short samples of conductor and a central solenoid (CS) model coil were tested. All of the toroidal field (TF) and poloidal field (PF) coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600 meters long length jacketing line for cable-in-conduit conductors, two winding machines, a set of vacuum pressure impregnation (VPI) equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.
{"title":"Progress of HT-7U superconducting tokamak","authors":"Weng Peide","doi":"10.1109/FUSION.2002.1027695","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027695","url":null,"abstract":"The HT-7U is a superconducting tokamak, which is constructing in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). The mission of the HT-7U project is to develop scientific and engineering issues on the steady state operation of advanced tokamak. The engineering design of the device is optimized. The R&D program is going on. Short samples of conductor and a central solenoid (CS) model coil were tested. All of the toroidal field (TF) and poloidal field (PF) coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600 meters long length jacketing line for cable-in-conduit conductors, two winding machines, a set of vacuum pressure impregnation (VPI) equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79596869","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027668
M. Youssef, M. Sawan, H. Khater
In this paper, we quantitatively assess the advantages offered by thick liquid wall (LW) concepts over conventional solid wall (SW) concepts in terms of the substantial reduction in radiation damage as well as activation to solid structure with subsequent reduction in radwaste volume and hazard. The conventional SW FW/blanket considered is made of Li/V with a peak neutron wall load of 5 MW/m/sup 2/ (3.5 MW/m/sup 2/ ave.) and is compared to a thick liquid lithium with a peak neutron wall load of 10 MW/m/sup 2/ (7 MW/m/sup 2/ ave.). "Fixed Radii" and "Fixed Fusion Power" configurations are considered. To have a consistent comparison, the two blankets were optimized first such that adequate tritium breeding ratio is obtained and the same level of magnet protection against radiation damage is reached.
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