Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218669
J. Scoville, R. L. La Haye
Careful measurements have been made of the error fields created by the 19 field-shaping coils on DIII-D. Spectral analysis has yielded a better understanding of the source of the toroidal and poloidal field errors existing on the tokamak. With this new information, a much better error-field correction coil (C-coil) has been designed which should be able to nearly eliminate the m=2, n=1 component of the error field without introducing other toroidal modes. By dividing the C-coil into segments which can be connected in different configurations, one will be able to program the phase and amplitude of a variety of perturbation fields for experimentation. Alternatively, the coil can be configured and programmed to provide the maximum reduction of error fields for routine plasma operations, enabling a wider stable operating parameter space for the tokamak. Details for the C-coil design and its harmonic spectrum are presented.<>
{"title":"Design of a coil to correct magnetic field errors on the DIII-D tokamak","authors":"J. Scoville, R. L. La Haye","doi":"10.1109/FUSION.1991.218669","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218669","url":null,"abstract":"Careful measurements have been made of the error fields created by the 19 field-shaping coils on DIII-D. Spectral analysis has yielded a better understanding of the source of the toroidal and poloidal field errors existing on the tokamak. With this new information, a much better error-field correction coil (C-coil) has been designed which should be able to nearly eliminate the m=2, n=1 component of the error field without introducing other toroidal modes. By dividing the C-coil into segments which can be connected in different configurations, one will be able to program the phase and amplitude of a variety of perturbation fields for experimentation. Alternatively, the coil can be configured and programmed to provide the maximum reduction of error fields for routine plasma operations, enabling a wider stable operating parameter space for the tokamak. Details for the C-coil design and its harmonic spectrum are presented.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"43 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114634432","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218648
J.P. Smith, C. Baxi, E. Reis, M. Schaffer, K. Schaubel, M. Menon
A cryocondensation pump will be installed in the baffle chamber of the DIII-D tokamak in the spring of 1992. The design is complete and fabrication of this pump is in progress. The purpose of the pump is to study plasma density control by pumping the divertor. The pump is toroidally continuous, approximately 10 m long, in the lower outer corner of the vacuum vessel interior. It consists of a 1-m/sup 2/ liquid-helium-cooled surface surrounded by a liquid-nitrogen-cooled shield to limit the heat load on the helium-cooled surface. The stainless steel liquid nitrogen shell has a copper coating on it to enhance thermal conductivity but the coating is broken to keep the toroidal electrical resistance high. The liquid-nitrogen-cooled surface is surrounded by a radiation/particle shield to prevent energetic particles from impacting and releasing condensed water molecules. The whole pump is supported off the water-cooled vacuum vessel wall. A testing program was used to develop coating techniques to enhance heat transfer and emissivity of the various surfaces. Fabrication tests were done to determine the best method of attaching the liquid nitrogen flow tubes to their shield surfaces. A prototype sector of the pump was built to verify fabrication and assembly techniques.<>
{"title":"The design and fabrication of a toroidally continuous cryocondensation pump for the DIII-D advanced divertor","authors":"J.P. Smith, C. Baxi, E. Reis, M. Schaffer, K. Schaubel, M. Menon","doi":"10.1109/FUSION.1991.218648","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218648","url":null,"abstract":"A cryocondensation pump will be installed in the baffle chamber of the DIII-D tokamak in the spring of 1992. The design is complete and fabrication of this pump is in progress. The purpose of the pump is to study plasma density control by pumping the divertor. The pump is toroidally continuous, approximately 10 m long, in the lower outer corner of the vacuum vessel interior. It consists of a 1-m/sup 2/ liquid-helium-cooled surface surrounded by a liquid-nitrogen-cooled shield to limit the heat load on the helium-cooled surface. The stainless steel liquid nitrogen shell has a copper coating on it to enhance thermal conductivity but the coating is broken to keep the toroidal electrical resistance high. The liquid-nitrogen-cooled surface is surrounded by a radiation/particle shield to prevent energetic particles from impacting and releasing condensed water molecules. The whole pump is supported off the water-cooled vacuum vessel wall. A testing program was used to develop coating techniques to enhance heat transfer and emissivity of the various surfaces. Fabrication tests were done to determine the best method of attaching the liquid nitrogen flow tubes to their shield surfaces. A prototype sector of the pump was built to verify fabrication and assembly techniques.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129054895","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218656
P. Gohil, K. Burrell, R. Groebner, J. Kim, W. Martin, E. McKee, R. Seraydarian
The charge exchange recombination, (CER) diagnostic system on the DIII-D tokamak is used to make spatially and temporally resolved measurements of the ion temperature and toroidal and poloidal rotation velocities. This is performed through visible spectroscopic measurements of the Doppler-broadened and Doppler-shifted He II 468.6-nm, CVI 529.1-nm, and BV 494.5-nm spectral lines which have been excited by charge exchange recombination interactions between the fully stripped ions and the neutral atoms from the heating beams. The plasma viewing optics comprises 32 viewing chords spanning a typical plasma minor radius of 63 cm across the midplane, of which 15 spatial chords span 4.2 cm at the plasma edge just within the separatrix and provide a chord-to-chord spatial resolution of 0.3 cm. Fast camera readout electronics can provide a temporal resolution of 260 mu s per time slice, but the effective minimum integration time, at present, is 1 ms, which is limited by the detected photon flux from the plasma and the decay times of the phosphors used on the multichannel plate image intensifiers. Significant changes in the edge plasma radial electric field at the L-H transition have been observed, as determined from the CER measurements, and these results are being extensively compared to theories which consider the effects of sheared electric fields on plasma turbulence.<>
{"title":"The charge exchange recombination diagnostic system on the DIII-D tokamak","authors":"P. Gohil, K. Burrell, R. Groebner, J. Kim, W. Martin, E. McKee, R. Seraydarian","doi":"10.1109/FUSION.1991.218656","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218656","url":null,"abstract":"The charge exchange recombination, (CER) diagnostic system on the DIII-D tokamak is used to make spatially and temporally resolved measurements of the ion temperature and toroidal and poloidal rotation velocities. This is performed through visible spectroscopic measurements of the Doppler-broadened and Doppler-shifted He II 468.6-nm, CVI 529.1-nm, and BV 494.5-nm spectral lines which have been excited by charge exchange recombination interactions between the fully stripped ions and the neutral atoms from the heating beams. The plasma viewing optics comprises 32 viewing chords spanning a typical plasma minor radius of 63 cm across the midplane, of which 15 spatial chords span 4.2 cm at the plasma edge just within the separatrix and provide a chord-to-chord spatial resolution of 0.3 cm. Fast camera readout electronics can provide a temporal resolution of 260 mu s per time slice, but the effective minimum integration time, at present, is 1 ms, which is limited by the detected photon flux from the plasma and the decay times of the phosphors used on the multichannel plate image intensifiers. Significant changes in the edge plasma radial electric field at the L-H transition have been observed, as determined from the CER measurements, and these results are being extensively compared to theories which consider the effects of sheared electric fields on plasma turbulence.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115857068","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218723
G. Bramson
The author describes recent modifications to the DIII-D central timing system hardware and software to allow more flexible and complex timing waveforms than previously available to control NB (neutral beam) injection. In the past, for each plasma discharge, the start time and the pulse duration for each NB source could be set. The new NB timing system allows the experimental session leader to specify more than one pulse for each NB source to make efficient use of multiple physics experiments in a single discharge. Also, pulse trains with programmable duty cycles can be independently specified for each NB source to make NB injection a more useful diagnostic tool. Experimental results using features of the new NB timing system are presented.<>
{"title":"Timing system for neutral beam injection on the DIII-D tokamak","authors":"G. Bramson","doi":"10.1109/FUSION.1991.218723","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218723","url":null,"abstract":"The author describes recent modifications to the DIII-D central timing system hardware and software to allow more flexible and complex timing waveforms than previously available to control NB (neutral beam) injection. In the past, for each plasma discharge, the start time and the pulse duration for each NB source could be set. The new NB timing system allows the experimental session leader to specify more than one pulse for each NB source to make efficient use of multiple physics experiments in a single discharge. Also, pulse trains with programmable duty cycles can be independently specified for each NB source to make NB injection a more useful diagnostic tool. Experimental results using features of the new NB timing system are presented.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"190 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121729358","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218831
E. Reis, R. Blevins, T. Jensen, J. Luxon, P. Petersen, E. Strait
The motions of the DIII-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which was input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hz that are sufficiently loud to be felt as well as heard by the DIII-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds.<>
{"title":"Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities","authors":"E. Reis, R. Blevins, T. Jensen, J. Luxon, P. Petersen, E. Strait","doi":"10.1109/FUSION.1991.218831","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218831","url":null,"abstract":"The motions of the DIII-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which was input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hz that are sufficiently loud to be felt as well as heard by the DIII-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"os-27 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127770702","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218733
P. Petersen, S. Miller
Operation of the DIII-D tokamak involves many groups which work on the various subsystems. To overview and speed the solution to trouble or problem areas that limit machine availability, a common trouble report system was established. The TROUBLE database automates the recording of trouble reports and eases analysis of problem areas. It contains information on equipment affected, description of problem, cause of problem, solution to problem, and machine downtime (if any). It was created using S1032 from Compuserve Data Technologies and runs on a VAX 8650. The data are used to find the major problem areas so they can be solved and improve the tokamak availability. The data are available to Idaho National Engineering Laboratory (INEL). It is using the data with data from other tokamaks to develop a fusion failure experience data collection. The authors' experience is that a few failures are often the cause of a major part of the downtime. They discuss these failures and the actions taken to correct them. The database will also be used to determine the preventive maintenance schedule for different components.<>
DIII-D托卡马克的操作涉及许多小组，他们在各个子系统上工作。为了概述和加速解决限制机器可用性的故障或问题区域，建立了一个通用故障报告系统。TROUBLE数据库可以自动记录故障报告并简化问题区域的分析。它包含受影响设备的信息、问题描述、问题原因、问题解决方案和机器停机时间(如果有的话)。它是使用Compuserve Data Technologies的S1032创建的，运行在VAX 8650上。这些数据用于发现主要问题区域，以便解决这些问题并提高托卡马克的可用性。这些数据可供爱达荷国家工程实验室(INEL)使用。它正在将这些数据与其他托卡马克的数据一起开发聚变失败经验数据集。作者的经验是，一些故障通常是停机时间的主要原因。他们讨论这些失败和采取的措施来纠正它们。该数据库还将用于确定不同部件的预防性维护计划。
{"title":"The DIII-D tokamak TROUBLE report database","authors":"P. Petersen, S. Miller","doi":"10.1109/FUSION.1991.218733","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218733","url":null,"abstract":"Operation of the DIII-D tokamak involves many groups which work on the various subsystems. To overview and speed the solution to trouble or problem areas that limit machine availability, a common trouble report system was established. The TROUBLE database automates the recording of trouble reports and eases analysis of problem areas. It contains information on equipment affected, description of problem, cause of problem, solution to problem, and machine downtime (if any). It was created using S1032 from Compuserve Data Technologies and runs on a VAX 8650. The data are used to find the major problem areas so they can be solved and improve the tokamak availability. The data are available to Idaho National Engineering Laboratory (INEL). It is using the data with data from other tokamaks to develop a fusion failure experience data collection. The authors' experience is that a few failures are often the cause of a major part of the downtime. They discuss these failures and the actions taken to correct them. The database will also be used to determine the preventive maintenance schedule for different components.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"24 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133829162","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218736
J. Ferron, A. Kellman, E. McKee, T. Osborne, P. Petrach, T. Taylor, J. Wight, E. Lazarus
An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Through use of a control algorithm which can be varied in real time this system is expected to provide more precise control of the DIII-D discharge shape than is possible in the analog control system presently in use. In addition to shape control the system will provide a platform for research on real-time optimization of discharge performance. A relatively simple design has been achieved through the use of a small number of very-high-speed digital processors coupled with high-speed data acquisition hardware. The frequency response is expected to be adequate to control the unstable vertical motion of highly elongated discharge shapes.<>
{"title":"An advanced plasma control system for the DIII-D tokamak","authors":"J. Ferron, A. Kellman, E. McKee, T. Osborne, P. Petrach, T. Taylor, J. Wight, E. Lazarus","doi":"10.1109/FUSION.1991.218736","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218736","url":null,"abstract":"An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Through use of a control algorithm which can be varied in real time this system is expected to provide more precise control of the DIII-D discharge shape than is possible in the analog control system presently in use. In addition to shape control the system will provide a platform for research on real-time optimization of discharge performance. A relatively simple design has been achieved through the use of a small number of very-high-speed digital processors coupled with high-speed data acquisition hardware. The frequency response is expected to be adequate to control the unstable vertical motion of highly elongated discharge shapes.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128636385","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218734
K. L. Greene
In 1990, MFITD and MFITPLAY, programs used by users when running fusion plasma experiments, were converted from User Interface Services (UIS) routines to DECs X Windows System using the DECWindows window manager. These modifications were required because of a move by Digital Equipment Corporation (DEC) to support X windows and phase out UIS. Due to the nature and purpose of each program, MFITD needed only simple graphics conversion while MFITPLAY was completely rewritten. The DECWindows version of MFITPLAY offers a number of improvements, such as a more intuitive user interface.<>
1990年，用户在运行聚变等离子体实验时使用的程序MFITD和MFITPLAY使用DECWindows窗口管理器从用户界面服务(UIS)例程转换为DECs X Windows系统。这些修改是必要的，因为数字设备公司(DEC)支持X窗口和逐步淘汰美国。由于每个程序的性质和目的，MFITD只需要简单的图形转换，而MFITPLAY则完全重写。DECWindows版本的MFITPLAY提供了许多改进，例如更直观的用户界面。
{"title":"Displaying DIII-D plasma data using DEC's X Window System","authors":"K. L. Greene","doi":"10.1109/FUSION.1991.218734","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218734","url":null,"abstract":"In 1990, MFITD and MFITPLAY, programs used by users when running fusion plasma experiments, were converted from User Interface Services (UIS) routines to DECs X Windows System using the DECWindows window manager. These modifications were required because of a move by Digital Equipment Corporation (DEC) to support X windows and phase out UIS. Due to the nature and purpose of each program, MFITD needed only simple graphics conversion while MFITPLAY was completely rewritten. The DECWindows version of MFITPLAY offers a number of improvements, such as a more intuitive user interface.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129840225","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218715
J.C. Allen, R. Callis, W. Cary, T. E. Harris, A. Nerem
The DIII-D tokamak facility is currently upgrading its electron cyclotron resonance heating system. The new system is based on 110-GHz gyrotrons developed by Varian. As part of this upgrade, the superconducting magnet power supplies were required to be remotely controlled and monitored accurately. The 110-GHz gyrotron superconducting magnet has eight coils, which are energized by current regulating power supplies. An analog-to-digital (A/D) system was designed to allow remote coil current monitoring and power supply programming. The A/D system is an eight channel multiplexed, 16-b, bidirectional, fiber optically linked, A/D telemetry system. Design concerns and tradeoffs are discussed, as are the results of in-system use.<>
{"title":"Eight channel-16 bit, bidirectional analog to digital monitoring and control system","authors":"J.C. Allen, R. Callis, W. Cary, T. E. Harris, A. Nerem","doi":"10.1109/FUSION.1991.218715","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218715","url":null,"abstract":"The DIII-D tokamak facility is currently upgrading its electron cyclotron resonance heating system. The new system is based on 110-GHz gyrotrons developed by Varian. As part of this upgrade, the superconducting magnet power supplies were required to be remotely controlled and monitored accurately. The 110-GHz gyrotron superconducting magnet has eight coils, which are energized by current regulating power supplies. An analog-to-digital (A/D) system was designed to allow remote coil current monitoring and power supply programming. The A/D system is an eight channel multiplexed, 16-b, bidirectional, fiber optically linked, A/D telemetry system. Design concerns and tradeoffs are discussed, as are the results of in-system use.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126296077","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}
Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218919
M. J. Schaffer, S. Lippmann, M. Mahdavi, T. Petrie, R. Stambaugh, J. Hogan, C. Klepper, P. Mioduszewski, L. W. Owen, D. N. Hill, M. Rensink, D. Buchenauer
A novel, electrically biasable, semiclosed divertor was installed and operated in the DIII-D lower outside divertor location. The semiclosed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. Electrical bias controls the distribution of particle recycle between the inner and outer divertors by E*B drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduces the sensitivity of plenum pressure to separatrix position. In particular, E*B drifts in the DIII-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density.<>
{"title":"Particle control in the DIII-D advanced divertor","authors":"M. J. Schaffer, S. Lippmann, M. Mahdavi, T. Petrie, R. Stambaugh, J. Hogan, C. Klepper, P. Mioduszewski, L. W. Owen, D. N. Hill, M. Rensink, D. Buchenauer","doi":"10.1109/FUSION.1991.218919","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218919","url":null,"abstract":"A novel, electrically biasable, semiclosed divertor was installed and operated in the DIII-D lower outside divertor location. The semiclosed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. Electrical bias controls the distribution of particle recycle between the inner and outer divertors by E*B drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduces the sensitivity of plenum pressure to separatrix position. In particular, E*B drifts in the DIII-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"147 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126188200","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}
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