2007 IEEE 22nd Symposium on Fusion Engineering最新文献

{"title":"Phase Lag Infra-red Thermal Examination (PLITE); A New Non-destructive Test Process","authors":"J. M. McDonald, T. Lutz, M. Ulrickson, T. Tanaka, D. Youchison, R. Nygren","doi":"10.1109/FUSION.2007.4337873","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337873","url":null,"abstract":"The International Organization of ITER (International Thermonuclear Experimental Reactor) specifies a requirement of 3 mm in diameter for the largest permissible flaw in the joint of the beryllium (Be) armor tiles and the underlying heat sink made of a copper-chrome-zirconium (CuCrZr) alloy for the first wall (FW). We investigated the sensitivity of a new non-destructive process of detecting these flaws using a method in which we mapped the phase lag of the temperatures on the surface of a sample during thermal cycling with a sinusoidally varying water temperature. A method with hot-cold water test that we had pioneered during the 1990's for the development of a water-cooled mid-plane modular limiter for Tore Supra had worked well with the high conductivity armor made of pyrolytic graphite brazed to copper tubes. The paper describes the experimental system, test samples and some experimental results.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130458993","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}

{"title":"Development Status of the Helium-Cooled Porous Tungsten Heat Exchanger Concept","authors":"S. Sharafat, A. Aoyama, M. Narula, J. El-Awady, N. Ghoniem, B. Williams, D. Youchison","doi":"10.1109/FUSION.2007.4337888","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337888","url":null,"abstract":"The development status of a helium cooled refractory metal heat exchanger (HX) concept using tungsten foam for enhanced heat transfer is presented. The HX design is based on azimuthal flow of helium through the foam sandwiched between two concentric tungsten tubes. This concept holds the promise for an efficient and low pressure-drop HX concept for plasma facing components, such as divertors. A prototypical flat-top HX-tube is being manufactured for testing at the high heat flux testing facility at SNL. Concept design optimization requires knowledge of the enhanced heat transfer coefficients due to the foam structure. Solid models of representative metal foams were developed for use in CFD analysis. Initial CFD results show improved heat transfer between the heated wall to the coolant. For a 1-mm thick foam with a specific density of 12% and a pore density of 65 PPI an average heat transfer coefficients of 40 000 W/m2-K was estimated, along with a pressure drop of ~60 kPa. For a 10 MW/m2 surface heat load and an inlet helium temperature of 600degC at a pressure of 4 MPa, maximum structural temperatures were estimated to be 1060degC. This preliminary design has a maximum combined primary plus secondary von Mises stress of less than 600 MPa.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"116 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133513170","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}

{"title":"High Pressure Supersonic Gas Jet Fueling on NSTX","authors":"V. Soukhanovskii, M. Bell, W. Blanchard, J. Dong, R. Gernhardt, R. Kaita, H. Kugel, T. Provost, A. Roquemore, P. Sichta","doi":"10.1109/FUSION.2007.4337860","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337860","url":null,"abstract":"A supersonic gas injector (SGI) has been developed for fueling and diagnostic applications on NSTX. The SGI is comprised of a small de Laval converging-diverging graphite nozzle, a commercial piezoelectric gas valve, and a diagnostic package, all mounted on a movable probe at a low field side midplane port location. The nozzle operated in a pulsed regime at room temperature, reservoir deuterium pressure up to 2500 Torr (50 PSIA), flow rate up to 65 Torr 1 /s (4.55e2f particles/s), and a measured Mach number of about 4. In initial experiments the SGI was used for fueling of ohmic and 2 -6 MW NBI-heated L-and H-mode plasmas. Reliable H-mode access was obtained with SGI fueling, with a fueling efficiency in the range 0.1 -0.3. Good progress was also made toward a controlled density SGI-fueled H-mode plasma scenario with the flow rate of the uncontrolled high field side (HFS) gas injector reduced by up to 20. These experiments motivated a number of SGI upgrades: 1) the maximum plenum pressure has been increased to 5000 Torr (100 PSIA), 2) the plenum pressure volume has been doubled, 3) the gas delivery system has been changed to allow for injection of various gases, 4) a multi-pulse capability has been implemented. As a result of the upgrades, the maximum flow rate increased to about 130 Torr 1 /s. Laboratory gas jet characterization tests indicated a Mach number of about 4 with H2 and I) , and 4-6 with He and N2. Plasma experiments demonstrated the high-pressure gas jet fueling compatibility with H-mode plasmas, high fueling efficiency (0.1 -0.3), and high SOL penetration.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130937035","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}

{"title":"Design status of the ITER ECH Upper Launcher Steering Mirror Mechanism","authors":"J. Landis, R. Chavan, M. Henderson, R. Bertizzolo, A. Collazos, F. Sanchez","doi":"10.1109/FUSION.2007.4337928","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337928","url":null,"abstract":"The ITER ECRH upper port antenna (or launcher) will be used to drive current locally for stabilising the neoclassical tearing mode (NTM) by depositing ram-wave power inside of the island which forms on the q=3/2 or 2 rational magnetic flux surfaces and control the sawtooth instability by driving current near the q=l surface. This requires the launcher to be capable of steering the focused beam deposition location across the resonant flux surface over the range where the q=l, 3/2 and 2 surfaces are expected to be found (roughly the outer half of the plasma). ITER'S present reference design uses a front steering (FS) concept, which uses a moveable mirror close to the plasma. Two separate mirrors are used to decouple the focusing and steering aspects resulting in an optimized optical configuration providing a well focused beam over a large steering range. The steering mechanism providing the mirror rotation uses a frictionless and backlash free mechanical system based on the elastically compliant deformation of structural components to avoid the in vessel tribological difficulties. Traditional designs are based on push-pull rods acting on a mirror which rotates with ball bearings, they present the risk of gripping or result in stick-slip movements. The ball bearings are replaced with a set of flexure pivots while the classic actuation through a push-pull rod scheme is replaced by a directly acting pneumatic system consisting on a fast feed line, bellows and springs, in which the pressure acting on the bellows pushes the mirror against the compressive springs. The rotation of the mirror is thus produced by the counteraction between the forces exerced by the springs and the bellows, themselves piloted by the pressure of the system. A servovalve placed outside of the port plug and connected to the bellows by a small tube will control this pressure. The system also includes flexible water cooling pipes which allow the removal of heat generated by the ohmic surface losses of the reflected mm-wave beams and the nuclear and radiation volumic heating of the rotating mirror components. This paper will give an overview of the engineering and design issues and their solutions, and provide the development status of the different components of the mechanism. Special attention will be given to the engineering analysis performed to ensure compliance of the steering mechanism with the various ITER requirements.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132787636","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}

{"title":"Hydrogen isotopes permeation evaluation in the advanced material for nuclear fusion blanket use","authors":"T. Minami, S. Niigawa, Y. Ueno, T. Hinoki, Y. Yamamoto, S. Konishi","doi":"10.1109/FUSION.2007.4337867","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337867","url":null,"abstract":"Behavior of hydrogen isotopes such as permeation, diffusion, and dissolution infusion blanket materials were investigated from a viewpoint of the material development and a material design for a fusion reactor. Experiments were conducted with Reduced Activation Ferritic Steel (F82H) and various kinds of SiC materials at elevated temperature. To evaluate permeability of SiC materials and RAFM, deuterium gas permeability was measured using newly designed device. Permeation of hydrogen through RAFM was not significantly different from that of Austenitic Steel, however the temperature dependence of the permeability diffusivity, and solubility showed marked discontinuous change around 850 degree C. Change in crystal structure from bcc to fee is a suspected cause. The measurement of permeability of deuterium gas in Hexoloy and CVD SiC samples were attempted at the temperature above 800 degree C The permeability and deuterium diffusivity of Hexoloy is 2 orders of magnitude smaller than that of CVD SiC. On the pressure dependence of permeability, both linear and square-root dependence were seen.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132647286","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}

{"title":"Pellet Dropper Device for ELM Control on DIII-D","authors":"S. Combs, L. Baylor, C. Foust, J. McGill, J. Caughman, D. Fehling, M. Hansink, T. Jernigan, D. Rasmussen","doi":"10.1109/FUSION.2007.4337861","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337861","url":null,"abstract":"On several experimental tokamaks, pellet injection has been found to trigger edge localized modes (ELMs) in H-mode plasmas. This can provide a technique for ELM amelioration by reducing the ELM size with small high-frequency pellets. The key for success appears to be small pellets that penetrate just beyond the separatix, enough to trigger an ELM, but not enough to strongly fuel the plasma. To provide a source of small pellets, a pellet dropper device has been developed at the Oak Ridge National Laboratory and installed on the DIII-D tokamak. The pellet dropper consists of a batch extruder with an exit nozzle to provide a filament of solid deuterium (nominal 1-mm diameter), from which pellets are punched/dropped at rates of up to ap50 Hz and at speeds of <10 m/s. The pellets are propelled directly downward and through a vertical injection port on DIII-D. In this paper, the design and the initial test results are presented, and the installation on DIII-D is described.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"150 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115778608","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}

{"title":"A Conceptual Design for the Magnets in an IFE Magnetic Intervention Chamber","authors":"T. Dodson, F. Dahlgren, I. Zatza, C. Gentile, C. Priniski, T. Kozub, G. Gettelfinger, J. Sethian, A. E. Robson","doi":"10.1109/FUSION.2007.4337877","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337877","url":null,"abstract":"A conceptual design for a magnetic intervention system is presented in support of a 2 GW IFE direct drive fusion power reactor. The system is designed employing a cusp field to deflect ions generated by an IFE implosion away from the first wall of the reactor core and into specifically designed ion dumps. The magnetic coil system will employ liquid helium cooled 5083 Aluminum alloy casing on a Rutherford NbTi cable. The cables are configured as four double pancakes with a 5083 Aluminum alloy case for structural support. The conceptual design and corresponding preliminary load and field calculations will be presented.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115805445","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}

{"title":"NCSX Component Fabrication Challenges","authors":"L. Dudek, J. Chrzanowski, M. Viola, P. Heitzenroeder, T. Meighan, S. Raftopoulos","doi":"10.1109/FUSION.2007.4337863","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337863","url":null,"abstract":"The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The stellarator core is designed to produce a compact 3-D plasma that combines stellarator and tokamak physics advantages. The complex geometry and tight fabrication tolerances of NCSX create some unique engineering and assembly challenges. This paper will describe a few of the challenges of the machine's Modular Coils and vacuum vessel field period assembly and how they are being solved. Coil assembly began in November 2005 and to date 3 Modular Coils have been completed. One vacuum vessel 120deg section has been delivered and field period assembly work began in May 2006. Machine sector sub-assembly, machine assembly, and testing will follow, leading to First Plasma in 2011.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"51 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125093335","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}

{"title":"The Coupler Protection System Upgrade for Lower Hybrid Current Drive on Alcator C-Mod","authors":"W. Burke, D. Terry, H. Kennedy, J. Stillerman, J. McLean, P. Milne","doi":"10.1109/FUSION.2007.4337937","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337937","url":null,"abstract":"The MIT Plasma Science and Fusion Center (PSFC) Alcator C-Mod project has implemented a lower hybrid current drive (LHCD) system that uses 12 VKC-7849 Varian (CPI) klystron transmitters to supply up to 3 MW source power at 4.6 GHz. Power from each transmitter is split eight ways using a complex system of waveguides to drive the 96-channel coupler array. The existing Coupler Protection System (CPS) monitors directional coupler forward and reflected power signals at 60 locations in the system and provides 60 channels of voltage standing wave ratio (VSWR) protection for the coupler array. To further improve coupler protection and provide remote adjustability an upgrade to the CPS is underway. The most efficient approach to this upgrade has been to use an existing fast digitizer board design having MDS+ data system communication capabilities and reserve field programmable gate array (FPGA) circuitry. Working closely with the digitizer board manufacturer and designers, D-TACQ Solutions, PSFC engineers are developing CPS fault protection logic for twelve each of the ten channel, 14 bit, 6 MSPS/channel digitizer boards with associated custom rear transition modules (RTM). These boards will provide the protection and data acquisition functions needed to allow programmable, optimized protection and monitoring for the coupler and accessibility to the existing MIT PSFC MDS+ data acquisition and control system. Details of the CPS upgrade system will be presented.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"8 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129137268","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}

{"title":"NSTX OH Coil Design Improvements","authors":"M. Kalish, J. Chrzanowski, C. Neumeyer, B. Paul, R. Woolley, C. Jun","doi":"10.1109/FUSION.2007.4337876","DOIUrl":"https://doi.org/10.1109/FUSION.2007.4337876","url":null,"abstract":"The National Spherical Torus Experiment (NSTX) has been operating successfully since February of 1999. A unique element of NSTX is the center solenoid or OH coil that from the start has been a design challenged by the low aspect ratio/geometry of the device. To achieve this low aspect ratio the OH coil's outer diameter is constrained to a narrow profile creating the need for creative design solutions concerning cooling connections, lead orientation, and insulation schemes. The original design has succeeded overall, but NSTX run time has been lost due to coil reliability issues. It was decided in the last year that it would be prudent to fabricate a new OH coil and have it available as an upgrade to the experiment. The experience of operating and maintaining the OH coil has provided the basis for an improved OH coil design. A collaboration was arranged with ASIPP in China to fabricate a spare coil for NSTX. The new OH coil will incorporate both design improvements intended to increase reliability as well as upgrades that will provide flexibility during future operation by allowing for an expanded operational profile. This paper summarizes and reviews these design and reliability improvements.","PeriodicalId":124369,"journal":{"name":"2007 IEEE 22nd Symposium on Fusion Engineering","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2007-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128776868","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}