15th IEEE/NPSS Symposium. Fusion Engineering最新文献_第9页

Start-up simulations of the PULSAR pulsed tokamak reactor PULSAR脉冲托卡马克反应堆的启动模拟

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518408

K. Werley, C. Bathke

引用次数: 1

Confinement capability of ITER-EDA design ITER-EDA设计约束能力

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518311

N. Uckan

{"title":"Confinement capability of ITER-EDA design","authors":"N. Uckan","doi":"10.1109/FUSION.1993.518311","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518311","url":null,"abstract":"Confinement capability of the ITER-EDA (R=7.75 m, I=25 MA) operational scenarios is evaluated and compared with the ITER CDA (R=6 m, 22 MA). The ignition capability of ITER EDA is somewhat higher than that of CDA by a factor of 1.1-1.2 with empirical power law scalings and by a factor of 1.5-2 with offset linear scalings. Simulations with the RLW /spl chi/(/spl nabla/T/sub e/)/sub crit/ model show that both the EDA and CDA scenarios operates in L-mode, however CDA ignition margin is much smaller. With empirical scalings, the required L-mode confinement enhancement factor [H=/spl tau//sub E///spl tau//sub E/(scaling)] corresponding to, for example, ITER89-P L-mode scaling would be 1.5-1.6 in ITER EDA relative to 1.8 in CDA for 10% He (plus 1% Be) concentration. At a higher concentration of He of 20-25%, the confinement capability is deteriorated and the required confinement enhancement factor (over empirical L-mode scalings) is /spl ges/2. The Ohmic confinement time is a factor of two higher in the EDA design (as compared to the CDA), yielding a strong reduction in the auxiliary power required to reach ignition. In 1.5-D simulations with L-mode enhancement factors of H/spl ges/1.2 allowed ohmic ignition with 25 MA, ignition was aided by initially peaked density profiles (and low He content) during the start-up.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"30 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117248900","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: 12

Design methods and actual performances of conductors for the superconducting coils of tokamaks 托卡马克超导线圈导体的设计方法及实际性能

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518417

B. Turck, D. Bessette, D. Ciazynski, J. Duchateau

引用次数: 9

Experiments with the CNPM gas gun for minimizing piston wear and propellant gas flow CNPM气枪减少活塞磨损和推进剂气体流量的实验

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518275

A. Reggiori, R. Carlevaro, G. Riva, G. Daminelli

引用次数: 2

Methodology for first wall design 第一面墙设计方法

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518532

J. Galambos, D. Conner, P. Goranson, D. Lousteau, D. Williamson, B. Nelson, F. Davis

引用次数: 0

Design of the TPX limiter and armor components TPX限位器及装甲部件的设计

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518536

D. Sevier, E. Chin, T.R. Hodapp, R. Junge, K. Redler, H. Mantz

{"title":"Design of the TPX limiter and armor components","authors":"D. Sevier, E. Chin, T.R. Hodapp, R. Junge, K. Redler, H. Mantz","doi":"10.1109/FUSION.1993.518536","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518536","url":null,"abstract":"The TPX limiter and armor systems are designed for steady-state operation from day one operation, at 18 MW plasma input power, to a possible upgrade to 45 MW. All components are designed for remote handling. Carbon-carbon (C-C) composites are the baseline plasma facing material for all limiter and armor systems. Where applicable, all components are made from low activation materials. The TPX limiter system consists of the inboard toroidal limiter, the outboard toroidal limiter, and three discrete poloidal limiters. These limiters are used for plasma startup and to protect the vessel, passive plates, and equipment in the ports from the energetic particle fluxes during steady-state operation. In addition, the inboard limiter protects the vacuum vessel from steady-state neutral beam shine-though and from neutral beam faults. The TPX armor components consist of two major systems: the neutral beam armor that protects the outer vessel wall and equipment in the ports, and the ripple armor that intercepts the trapped energetic particles that are drifting vertically in the ripple region. Different design concepts are employed for these plasma facing components (PFC) depending on their expected heat loads. Inboard and outboard limiters are designed with mechanically restrained C-C composite tiles mounted on cooled support plates. Components which must withstand higher heat loads, such as neutral beam and ripple armor, are made of C-C composite tiles brazed to actively-cooled copper.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"151 ","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114091637","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: 3

Argon frost continuous cryopump for fusion applications 用于聚变应用的氩冻连续低温泵

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518500

C. A. Foster, H. C. McCurdy

{"title":"Argon frost continuous cryopump for fusion applications","authors":"C. A. Foster, H. C. McCurdy","doi":"10.1109/FUSION.1993.518500","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518500","url":null,"abstract":"A cryopumping system based on the snail continuous cryopump concept is being developed for fusion applications under a DOE SBIR grant. The primary pump is a liquid helium cooled compound pump designed to continuously pump and fractionate deuterium/tritium and helium. The D/T pumping stage is a 500 mm bore cryocondensation pump with a nominal pumping speed of 45,000 L/s. It will be continuously regenerated by a snail regeneration head every 12 minutes. Continuous regeneration will dramatically reduce the vulnerable tritium inventory in a fusion reactor. Operating at an inlet pressure of 1 millitorr, eight of these pumps could pump the projected D/T flow in the ITER CDA design while reducing the inventory of tritium in the pumping system from 630 to 43 grams. The helium fraction will be pumped in a compound argon frost stage. This stage will also operate continuously with a snail regeneration head. In addition the argon spray head will be enclosed inside the snail, thereby removing gaseous argon from the process chamber. Since the cryocondensation stage will intercept over 90% of the D/T/H stream, a purified stream from this stage could be directly reinjected into the plasma as gas or pellets, thereby bypassing the isotope separation system and further simplifying the fuel cycle. Experiments were undertaken in Phase I which demonstrated continuous cryosorption pumping of hydrogen on CO/sub 2/ and argon frosts. The pumping system and its relevance to fusion reactor pumping will be discussed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114232649","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: 4

Drastic improvement of I/sub c/ of Nb/sub 3/Sn CIC conductor by prestraining at room temperature 室温下抑制对Nb/sub 3/Sn CIC导体I/sub c/性能的显著改善

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518530

A. Torossian, W. Specking, J. Duchateau, P. Decool

引用次数: 0

Tritium effects on the performance of turbomolecular pumps 氚对涡轮分子泵性能的影响

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518286

A. G. Heics, T. Shmayda, D. Muller

{"title":"Tritium effects on the performance of turbomolecular pumps","authors":"A. G. Heics, T. Shmayda, D. Muller","doi":"10.1109/FUSION.1993.518286","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518286","url":null,"abstract":"A special 150 L/s turbomolecular pump was used to pump pure tritium in closed loop mode over the course of six separate tritium runs. About 5.3 kCi of tritium was pumped during a cumulative period of 237 days. The pump, supplied by Leybold Heraeus, Germany for the purpose of tritium testing, features heavy duty bearings and motor, metal body seals, and radiation resistant coatings on electrical components. No tritium related degradation of the pump has been observed after 12,000 hours of operation. A liquid nitrogen cooled trap, used to trap the outgassing products from the system, required periodic regeneration. The loop was vacuum baked and recharged with tritium every 14 to 63 days of tritium testing. A bearing failure at 7176 hours is believed to have resulted from excessive bearing loads imposed by high system pressures. Testing was resumed after the original oil lubricated steel bearings were replaced with grease lubricated ceramic bearings. Vibration monitoring of turbomolecular pump bearings can be effectively used to predict remaining bearing life and to schedule maintenance. The study is intended to continue until the pump has undergone one year of cumulative tritium service.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"31 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114751101","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

Modelling of quench in CICC with a central channel in the conduit 管道中有中心通道的中金淬火模型

15th IEEE/NPSS Symposium. Fusion Engineering Pub Date : 1993-10-11 DOI: 10.1109/FUSION.1993.518524

J. Freidberg, A. Shajii, E. Chaniotakis, J. Mccarrick

引用次数: 2