Pub Date : 1991-11-01DOI: 10.1109/FUSION.1991.218701
W. Cary, J.C. Allen, R. Callis, J. Doane, T. E. Harris, C. Moeller, A. Nerem, R. Prater, D. Remsen
A new high-power electron cyclotron heating (ECH) system has been introduced on DIII-D. This system is designed to operate at 110 GHz with a total output power of 2 MW. The system consists of four Varian VGT-8011 gyrotrons, (output power of 500 kW), and their associated support equipment. All components have been designed for up to a 10-s pulse duration. The 110-GHz system is intended to further progress in RF current drive experiments on DIII-D when used in conjunction with the existing 60-GHz ECH (1.6-MW) and 30-60-MHz ICH (2-MW) systems. H-mode physics, plasma stabilization experiments, and transport studies are also to be conducted at 110 GHz. The system design philosophy was based on experience gained from the existing 60-GHz ECH system. The consequences of these design decisions are addressed, as is the performance of various 110-GHz components.<>
{"title":"110 GHz ECH on DIII-D: system overview and initial operation","authors":"W. Cary, J.C. Allen, R. Callis, J. Doane, T. E. Harris, C. Moeller, A. Nerem, R. Prater, D. Remsen","doi":"10.1109/FUSION.1991.218701","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218701","url":null,"abstract":"A new high-power electron cyclotron heating (ECH) system has been introduced on DIII-D. This system is designed to operate at 110 GHz with a total output power of 2 MW. The system consists of four Varian VGT-8011 gyrotrons, (output power of 500 kW), and their associated support equipment. All components have been designed for up to a 10-s pulse duration. The 110-GHz system is intended to further progress in RF current drive experiments on DIII-D when used in conjunction with the existing 60-GHz ECH (1.6-MW) and 30-60-MHz ICH (2-MW) systems. H-mode physics, plasma stabilization experiments, and transport studies are also to be conducted at 110 GHz. The system design philosophy was based on experience gained from the existing 60-GHz ECH system. The consequences of these design decisions are addressed, as is the performance of various 110-GHz components.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"21 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":"128285272","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.1991.218650
K. Schaubel, C. Baxi, G. Campbell, A. M. Gootgeld, A. Langhorn, G. J. Laughon, J. P. Smith, P. M. Anderson, M. M. Menon
The design of the cryogenic system for the DIII-D advanced divertor cryocondensation pump is presented. The advanced divertor incorporates a baffle chamber and bias ring located near the bottom of the DIII-D vacuum vessel. A 50000-l/s cryocondensation pump will be installed underneath the baffle for plasma particle exhaust. The pump consists of a liquid-helium-cooled tube operating at 4.3 K and a liquid-nitrogen-cooled radiation shield. Liquid helium is fed by forced flow through the cryopump. Compressed helium gas flowing through the high-pressure side of a heat exchanger is regeneratively cooled by the two-phase helium leaving the pump. The cooled high-pressure gaseous helium is then liquefied by a Joule-Thomson expansion valve. The liquid is returned to a storage dewar. The liquid nitrogen for the radiation shield is supplied by forced flow from a bulk storage system. Control of the cryogenic system is accomplished by a programmable logic controller.<>
{"title":"Design of the advanced divertor pump cryogenic system for DIII-D","authors":"K. Schaubel, C. Baxi, G. Campbell, A. M. Gootgeld, A. Langhorn, G. J. Laughon, J. P. Smith, P. M. Anderson, M. M. Menon","doi":"10.1109/FUSION.1991.218650","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218650","url":null,"abstract":"The design of the cryogenic system for the DIII-D advanced divertor cryocondensation pump is presented. The advanced divertor incorporates a baffle chamber and bias ring located near the bottom of the DIII-D vacuum vessel. A 50000-l/s cryocondensation pump will be installed underneath the baffle for plasma particle exhaust. The pump consists of a liquid-helium-cooled tube operating at 4.3 K and a liquid-nitrogen-cooled radiation shield. Liquid helium is fed by forced flow through the cryopump. Compressed helium gas flowing through the high-pressure side of a heat exchanger is regeneratively cooled by the two-phase helium leaving the pump. The cooled high-pressure gaseous helium is then liquefied by a Joule-Thomson expansion valve. The liquid is returned to a storage dewar. The liquid nitrogen for the radiation shield is supplied by forced flow from a bulk storage system. Control of the cryogenic system is accomplished by a programmable logic controller.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"253 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":"133838414","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.1991.218713
B. McHarg
The author describes the configuration of the DIII-D mixed interconnect VAX cluster and how it has improved the efficiency of data collection, system management, and personnel efficiency. The use of a VAX console system for central console management is also described. To improve the efficiency of the overall data acquisition system, a mixed interconnect VAX cluster has been formed consisting of 16 VAX computers. In the cluster, the software protocol for passing data around the cluster is much more efficient than using DECnet. The cluster has also greatly simplified the procedure of backing up disks. Another big improvement is the use of a VAX console system which ties all the console ports of the computers into one central computer system, which then manages the entire cluster.<>
{"title":"The use of a VAX cluster for the DIII-D data acquisition system","authors":"B. McHarg","doi":"10.1109/FUSION.1991.218713","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218713","url":null,"abstract":"The author describes the configuration of the DIII-D mixed interconnect VAX cluster and how it has improved the efficiency of data collection, system management, and personnel efficiency. The use of a VAX console system for central console management is also described. To improve the efficiency of the overall data acquisition system, a mixed interconnect VAX cluster has been formed consisting of 16 VAX computers. In the cluster, the software protocol for passing data around the cluster is much more efficient than using DECnet. The cluster has also greatly simplified the procedure of backing up disks. Another big improvement is the use of a VAX console system which ties all the console ports of the computers into one central computer system, which then manages the entire cluster.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"100 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":"130404304","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.1991.218651
C. Baxi, W. Obert
Thermal analysis of the nitrogen shield of the JET (Joint European Torus) cryopump was done using a finite element computer program. In this analysis, a parallel flow arrangement and two series flow arrangements were compared for cooldown from 300 to about 80 K. In order to simplify the analysis, coolant was assumed to be a N/sub 2/ gas at an inlet temperature of 80 K. It is shown that all three flow arrangements have similar time for cooling down the shield from 300 to 80 K. This means that the heat exchange effect or radial conduction from the warm part of the shield to the cold part of the shield for series flow arrangements is not dominant. Due to small conduction effects, it will be feasible to modify the design to a more stable series flow arrangement. This flow arrangement will also have minimum cooling time. The inner stainless steel shield has small thermal conductivity and, hence, this part of the shield lags in cooling behind the rest of the shield. This could be remedied by adding about a 1-mm layer of copper in poloidal stripes to the stainless steel fin.<>
利用有限元程序对JET (Joint European Torus)低温泵的氮屏蔽层进行了热分析。在此分析中，比较了平行流动安排和两种串联流动安排的冷却时间从300到大约80 K。为了简化分析，假设冷却剂为入口温度为80k时的N/sub / gas。结果表明，三种流动方式对300 ~ 80k的屏蔽层冷却时间相似。这意味着在串联流动布置中，从护板的温暖部分到护板的寒冷部分的热交换效应或径向传导不占主导地位。由于传导效应小，将设计修改为更稳定的串联流动布置是可行的。这种流动安排也将有最小的冷却时间。内部不锈钢护罩的导热系数很小，因此，这部分护罩的冷却滞后于护罩的其余部分。这可以通过在不锈钢翅片上加一层1毫米厚的极向条纹铜来弥补
{"title":"Optimization of thermal design for nitrogen shield of JET cryopump","authors":"C. Baxi, W. Obert","doi":"10.1109/FUSION.1991.218651","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218651","url":null,"abstract":"Thermal analysis of the nitrogen shield of the JET (Joint European Torus) cryopump was done using a finite element computer program. In this analysis, a parallel flow arrangement and two series flow arrangements were compared for cooldown from 300 to about 80 K. In order to simplify the analysis, coolant was assumed to be a N/sub 2/ gas at an inlet temperature of 80 K. It is shown that all three flow arrangements have similar time for cooling down the shield from 300 to 80 K. This means that the heat exchange effect or radial conduction from the warm part of the shield to the cold part of the shield for series flow arrangements is not dominant. Due to small conduction effects, it will be feasible to modify the design to a more stable series flow arrangement. This flow arrangement will also have minimum cooling time. The inner stainless steel shield has small thermal conductivity and, hence, this part of the shield lags in cooling behind the rest of the shield. This could be remedied by adding about a 1-mm layer of copper in poloidal stripes to the stainless steel fin.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"60 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":"127129863","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.1991.218845
P. Taylor
The DIII-D tokamak operates with a neutron radiation shield to allow enhanced plasma operations with increased neutron production while minimizing the site boundary dose level. Neutron rates as high as 4*10/sup 15/ neutron/s and total neutron production as high as 4*10/sup 15/ neutrons per shot are obtained while maintaining the site dose below the DOE (Department of Energy) administrative level of 20 mrem per year. The radiation shielding has increased by a factor of 300 over the preshield value and is in agreement with the design calculation. The maximum site neutron dose since installation of the shield has been less than 0.03 mrem for a shot and less than 0.4 mrem for a day. The site neutron and gamma dose are monitored continuously during operations by a PC-based computer system that provides the means of measuring the low dose levels that occur during a shot by including postshot background subtraction. The neutron and gamma dose are measured and archived by shot, hour, and day in a database. Activation of the machine after a run day and during vessel entries is monitored, and the activated nuclides have been determined.<>
{"title":"DIII-D radiation shielding procedures and experiences","authors":"P. Taylor","doi":"10.1109/FUSION.1991.218845","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218845","url":null,"abstract":"The DIII-D tokamak operates with a neutron radiation shield to allow enhanced plasma operations with increased neutron production while minimizing the site boundary dose level. Neutron rates as high as 4*10/sup 15/ neutron/s and total neutron production as high as 4*10/sup 15/ neutrons per shot are obtained while maintaining the site dose below the DOE (Department of Energy) administrative level of 20 mrem per year. The radiation shielding has increased by a factor of 300 over the preshield value and is in agreement with the design calculation. The maximum site neutron dose since installation of the shield has been less than 0.03 mrem for a shot and less than 0.4 mrem for a day. The site neutron and gamma dose are monitored continuously during operations by a PC-based computer system that provides the means of measuring the low dose levels that occur during a shot by including postshot background subtraction. The neutron and gamma dose are measured and archived by shot, hour, and day in a database. Activation of the machine after a run day and during vessel entries is monitored, and the activated nuclides have been determined.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"35 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":"122346392","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.1991.218719
J. Cummings, P. Thurgood
The Neural Beam Software Upgrade project was launched in early 1990. The major goals were to upgrade the MAX IV operating system to the latest version (K.1), use standard MODCOMP software (as much as possible), and to develop a very user-friendly, versatile system. Accomplishing these goals required new software to be developed and modifications to existing applications software to make it compatible with the latest operating system. The custom operating system modules to handle the message service and interrupt handling were replaced by the standard MODCOMP intertask communication (ITC) and interrupt routines that are part of the MAX IV operating system. The message service provides the mechanism for doing shot task sequencing. The interrupt routines are used to connect external interrupts to the system. The new software developed consists of a task dispatcher, a screen manager, and interrupt tasks. The existing applications software had to be modified to be compatible with the MODCOMP ITC services and consists of the Modcomp Infinity Data Base Manager, a multiuser system, and menu-driven operator system interface routines using the Infinity Data Base Manager.<>
{"title":"Software upgrade for the DIII-D neutral beam control systems","authors":"J. Cummings, P. Thurgood","doi":"10.1109/FUSION.1991.218719","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218719","url":null,"abstract":"The Neural Beam Software Upgrade project was launched in early 1990. The major goals were to upgrade the MAX IV operating system to the latest version (K.1), use standard MODCOMP software (as much as possible), and to develop a very user-friendly, versatile system. Accomplishing these goals required new software to be developed and modifications to existing applications software to make it compatible with the latest operating system. The custom operating system modules to handle the message service and interrupt handling were replaced by the standard MODCOMP intertask communication (ITC) and interrupt routines that are part of the MAX IV operating system. The message service provides the mechanism for doing shot task sequencing. The interrupt routines are used to connect external interrupts to the system. The new software developed consists of a task dispatcher, a screen manager, and interrupt tasks. The existing applications software had to be modified to be compatible with the MODCOMP ITC services and consists of the Modcomp Infinity Data Base Manager, a multiuser system, and menu-driven operator system interface routines using the Infinity Data Base Manager.<<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":"121455604","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.1991.218938
R. Pinsker, M. Mayberry, C. Petty, W. Cary, J. Pusl, D. Remsen, F. Baity, R. Goulding, D.J. Hofmann
The 2-MW fast wave current drive system on DIII-D is intended to provide near-term demonstration of up to 0.3 MA of current driven by the fast wave. The system used to drive the four-element phased antenna array which produces the required directional spectrum is presented. This system must be able to cope with strong coupling between antenna elements and the time-varying plasma load seen by the antennas. Computer modeling shows that this system should be able to maintain a directional spectrum at full power under most anticipated load conditions.<>
{"title":"30-60 MHz FWCD system on DIII-D: power division, phase control and tuning for a four-event antenna array","authors":"R. Pinsker, M. Mayberry, C. Petty, W. Cary, J. Pusl, D. Remsen, F. Baity, R. Goulding, D.J. Hofmann","doi":"10.1109/FUSION.1991.218938","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218938","url":null,"abstract":"The 2-MW fast wave current drive system on DIII-D is intended to provide near-term demonstration of up to 0.3 MA of current driven by the fast wave. The system used to drive the four-element phased antenna array which produces the required directional spectrum is presented. This system must be able to cope with strong coupling between antenna elements and the time-varying plasma load seen by the antennas. Computer modeling shows that this system should be able to maintain a directional spectrum at full power under most anticipated load conditions.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"9 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":"126870923","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 : 1991-10-03DOI: 10.1109/FUSION.1991.218911
W. Hogan
ICF (inertial confinement fusion) target design studies done for the Nova Upgrade have identified conditions under which the target ignition 'cliff' is shifted to much lower drive energy, albeit with the penalty that the gain achieved at a given drive energy is also smaller. These targets would repeatably produce the output and spectra of higher gain targets at low yield. They should thus allow building much smaller R&D reactors with full thermonuclear effects. Demonstration reactors at the 1 to 100 MW/sub e/ level appear to be feasible with driver energies of 0.5 to 2.0 MJ/pulse. These smaller, less expensive test and demonstration facilities should result in a lower IFE development cost. If the US government builds a driver and target factory, it is also conceivable that commercial organizations could build their own scaled concepts of IFE reactors using the beams and targets supplied by the government's facilities.<>
{"title":"Small inertial fusion energy (IFE) demonstration reactors","authors":"W. Hogan","doi":"10.1109/FUSION.1991.218911","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218911","url":null,"abstract":"ICF (inertial confinement fusion) target design studies done for the Nova Upgrade have identified conditions under which the target ignition 'cliff' is shifted to much lower drive energy, albeit with the penalty that the gain achieved at a given drive energy is also smaller. These targets would repeatably produce the output and spectra of higher gain targets at low yield. They should thus allow building much smaller R&D reactors with full thermonuclear effects. Demonstration reactors at the 1 to 100 MW/sub e/ level appear to be feasible with driver energies of 0.5 to 2.0 MJ/pulse. These smaller, less expensive test and demonstration facilities should result in a lower IFE development cost. If the US government builds a driver and target factory, it is also conceivable that commercial organizations could build their own scaled concepts of IFE reactors using the beams and targets supplied by the government's facilities.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129124788","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 : 1991-10-02DOI: 10.1109/FUSION.1991.218705
J. Woodworth, L. Chase, M. Guinan, W. Krupke, W.R. Sooy
A promising concept for the final optics of a baseline laser-driven ICF (inertial confinement fusion) reactor is described. It addresses the problem of long-term survival of the final optics with respect to neutron damage. The use of refractive optics is considered. A baseline design consists of two wedges of fused silica, which put a dogleg into the beam and thus remove optics further upstream from direct sight of the reactor. If the closest optic were located 40 m from the center of a 3-GWt reactor, it would be subject to an average 14-MeV neutron flux of approximately=5*10/sup 12/ n/s-cm/sup 2/ with a peak flux of approximately=6*10/sup 18/ n/s-cm/sup 2/. A major question to be answered is: What duration of reactor operation can this optic withstand? To answer this question, the literature bearing on radiation-induced optical damage in fused silica was reviewed to assess its implications for reactor operation with the baseline final optics scheme. It appears possible to continuously anneal the neutron damage in the silica by keeping the wedge at a modestly elevated temperature. The literature review indicates that the proposed final optic material-fused silica-is structurally resistant to radiation damage and that if operated at a temperature of about 300 degrees C will remain optically clear.<>
{"title":"Final optics for laser-driven inertial fusion reactors","authors":"J. Woodworth, L. Chase, M. Guinan, W. Krupke, W.R. Sooy","doi":"10.1109/FUSION.1991.218705","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218705","url":null,"abstract":"A promising concept for the final optics of a baseline laser-driven ICF (inertial confinement fusion) reactor is described. It addresses the problem of long-term survival of the final optics with respect to neutron damage. The use of refractive optics is considered. A baseline design consists of two wedges of fused silica, which put a dogleg into the beam and thus remove optics further upstream from direct sight of the reactor. If the closest optic were located 40 m from the center of a 3-GWt reactor, it would be subject to an average 14-MeV neutron flux of approximately=5*10/sup 12/ n/s-cm/sup 2/ with a peak flux of approximately=6*10/sup 18/ n/s-cm/sup 2/. A major question to be answered is: What duration of reactor operation can this optic withstand? To answer this question, the literature bearing on radiation-induced optical damage in fused silica was reviewed to assess its implications for reactor operation with the baseline final optics scheme. It appears possible to continuously anneal the neutron damage in the silica by keeping the wedge at a modestly elevated temperature. The literature review indicates that the proposed final optic material-fused silica-is structurally resistant to radiation damage and that if operated at a temperature of about 300 degrees C will remain optically clear.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"225 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-10-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127199262","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 : 1991-10-01DOI: 10.1109/FUSION.1991.218649
C. Baxi, G.J. Loughon, A. Langhorn, K. Schaubel, J.P. Smith, A. M. Gootgeld, G. Campbell, M. Menon
It is planned to install a cryogenic pump, in the lower divertor portion of the DIII-D tokamak with a pumping speed of 50000 l/s and an exhaust of 2670 Pa-l/s (20 torr-l/s). A coaxial counterflow configuration has been chosen for the helium panel of this cryogenic pump. The cool-down rates and fluid stability of this configuration are evaluated. A prototypic test was performed to increase confidence in the design. That the helium panel cooldown rate agreed quite well with analytical prediction and was within acceptable limits. The design flow rate proved stable and two-phase pressure drop can be predicted quite accurately. Results confirm that a cooldown of the helium panel from 300 K to liquid helium temperature can be achieved in the few minutes available between plasma shots. Helium flow results indicate that, at the design flow rate of 5 g/s, flow will be stable for heat loads up to 54 W. The expected heat load on the helium panel during operation is about 10 W.<>
{"title":"Verification test for helium panel of cryopump for DIII-D advanced divertor","authors":"C. Baxi, G.J. Loughon, A. Langhorn, K. Schaubel, J.P. Smith, A. M. Gootgeld, G. Campbell, M. Menon","doi":"10.1109/FUSION.1991.218649","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218649","url":null,"abstract":"It is planned to install a cryogenic pump, in the lower divertor portion of the DIII-D tokamak with a pumping speed of 50000 l/s and an exhaust of 2670 Pa-l/s (20 torr-l/s). A coaxial counterflow configuration has been chosen for the helium panel of this cryogenic pump. The cool-down rates and fluid stability of this configuration are evaluated. A prototypic test was performed to increase confidence in the design. That the helium panel cooldown rate agreed quite well with analytical prediction and was within acceptable limits. The design flow rate proved stable and two-phase pressure drop can be predicted quite accurately. Results confirm that a cooldown of the helium panel from 300 K to liquid helium temperature can be achieved in the few minutes available between plasma shots. Helium flow results indicate that, at the design flow rate of 5 g/s, flow will be stable for heat loads up to 54 W. The expected heat load on the helium panel during operation is about 10 W.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"30 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124023639","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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