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Sample records for diii-d vacuum vessel

  1. Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities

    International Nuclear Information System (INIS)

    Reis, E.; Blevins, R.D.; Jensen, T.H.; Luxon, J.L.; Petersen, P.I.; Strait, E.J.

    1991-11-01

    The motions of the D3-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 were 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 Hertz that are sufficiently loud to be felt as well as heard by the D3-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. The correlation of the theoretical and measured vessel currents, the dynamic measurements and analysis, and the acoustic measurements and analysis show that: (1) The physics model can predict vessel forces for selected values of plasma resistivity. The model also predicts poloidal and toroidal wall currents which agree with measured values; (2) The force-time history from the above model, used in conjunction with an axisymmetric structural model of the vessel, predicts vessel motions which agree well with measured values; (3) The above results, input to a simple acoustic model predicts the magnitude of sounds emitted from the vessel during disruptions which agree with acoustic measurements; (4) Correlation of measured vessel motions with structural analysis shows that a maximum vertical motion of the vessel up to 0.24 in will not overstress the vessel or its supports. 11 refs., 10 figs., 1 tab

  2. Design of the vacuum control system for DIII-D

    International Nuclear Information System (INIS)

    Campbell, G.L.; Callis, R.W.; Haskovec, J.S.; Heckman, E.J.; Moore, C.D.; Scoville, J.T.

    1986-01-01

    The vacuum control and instrumentation for the DIII-D upgrade was designed using a new large programmable controller with color graphic operator interfaces and intelligent distributed devices. Remote, optically isolated input and output is used as well as optical isolation for the operator and programming consoles. Gate valves between experimental equipment and the vacuum vessel are interlocked for machine safety by an intelligent interface based upon a commercially available microcontroller card. Complete automatic operation with capability for remote operator intervention was one goal of this design effort. The design of the system with emphasis on the graphics, optical isolation and microcontroller implementation will be discussed

  3. Optimized Baking of the DIII-D Vessel

    International Nuclear Information System (INIS)

    Anderson, P.M.; Kellman, A.G.

    1999-01-01

    The DIII-D tokamak vacuum vessel baking system is used to heat the vessel walls and internal hardware to an average temperature of 350 C to allow rapid conditioning of the vacuum surfaces. The system combines inductive heating and a circulating hot air system to provide rapid heating with temperature uniformity required by stress considerations. In recent years, the time to reach 350 C had increased from 9 hrs to 14 hrs. To understand and remedy this sluggish heating rate, an evaluation of the baking system was recently performed. The evaluation indicated that the mass of additional in-vessel hardware (50% increase in mass) was primarily responsible. This paper reports on this analysis and the results of the addition of an electric air heater and procedural changes that have been implemented. Preliminary results indicate that the time to 350 C has been decreased to 4.5 hours and the temperature uniformity has improved

  4. An interior vessel viewing system for DIII-D

    International Nuclear Information System (INIS)

    Senior, R.

    1989-11-01

    It was anticipated that there could be damage to the interior walls of the vacuum vessel during operations of the DIII-D tokamak. A method of viewing the inside of the vessel from the outside was required, that would allow the interior walls to be inspected visually for damage and to locate any debris resulting from operations. A miniature closed circuit television color camera system was developed which could be inserted into one of several ports of the vessel during a 'clean' vent, i.e., vented to inert gas. The system has pan, tilt and zoom capability and carries its own lighting. The use of this system allows a quick assessment of the condition of the vessel to be made under 'clean' vent conditions. This precludes the need for the permit process and manned entry into the vessel which would allow air inside the vessel. A permanent record of the inspection can then be made on video tape. The design and configuration of this camera system is presented and its use as a diagnostic tool discussed. 2 refs., 5 figs

  5. DIII-D in-vessel port cover and shutter assembly for the phase contrast interferometer

    International Nuclear Information System (INIS)

    Phelps, R.D.

    1994-01-01

    The entire outer wall of the DIII-D vacuum vessel interion is covered with a regular array of graphite tiles. Certain of the diagnostic ports through the outer vessel wall contain equipment which is shielded from the plasma by installing port covers designed to withstand energy deposition. If the diagnostic contained in the port must communicate with the vessel volume, a shutter assembly is usually provided. In the ports at 285 degrees, R+1 and R-1, interferometer mirrors have been installed to provide a means for transmitting a large diameter CO-2 laser beam through the edge of the plasma. To protect the mirrors and other hardware contained in these ports, a special protective plate and shutter arrangement has been designed. This report describes the details of design, fabrication, and installation of these protective covers and shutters

  6. The development of an in-vessel cryopump system for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Laughon, G.J.; Mahdavi, M.A.; Makariou, C.C.; Smith, J.P.; Schaffer, M.J.; Menon, M.M.

    1993-07-01

    The design, testing and initial operation of the DIII-D advanced divertor cryocondensation pumping system is presented. The pump resides inside the tokamak plasma containment vessel where it provides particle exhaust pumping, and it is subjected to Joule heating and hot particle heat loads during each 10 second discharge. In addition, the pump must withstand plasma disruption induced electromagnetic forces and 400 degrees C bake-out temperatures. Cooling is accomplished by forced flow liquid helium with the two-phase helium exhaust passing through a reliquefier for thermal efficiency. A prototype pump was constructed to study surface temperature rise as a function of flow geometry, applied heat load, helium mass flow rate, and pump outlet conditions. Prototype testing led to the development of a special geometry which was demonstrated to enhance two-phase flow stability and overall heat transfer. During initial operation, deuterium pumping speeds of 32,000 L/s at 2 mTorr pressure were achieved with a helium flow rate of 5 g/s. This speed was maintained during 300 W, 8 s long test beat pulses which meets operational goals

  7. Design of DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thurston, G.

    1989-11-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffling. The Advanced Divertor has two principal components: a toroidally symmetric baffle; and a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. All the feeds are supported from and maintain a 5 kV isolation to the vessel wall. 2 refs., 4 figs

  8. Design of DIII-D advanced divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thruston, G.

    1989-01-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffing. The Advanced Divertor has two principal components: ( 1) a toroidally symmetric baffle; and (2) a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. 2 refs., 4 figs

  9. Vacuum distilling vessel

    Energy Technology Data Exchange (ETDEWEB)

    Reik, H

    1928-12-27

    Vacuum distilling vessel for mineral oil and the like, characterized by the ring-form or polyconal stiffeners arranged inside, suitably eccentric to the casing, being held at a distance from the casing by connecting members of such a height that in the resulting space if necessary can be arranged vapor-distributing pipes and a complete removal of the residue is possible.

  10. Particle exhaust scheme using an in-vessel cryocondensation pump in the advanced divertor configuration of the DIII-D tokamak

    International Nuclear Information System (INIS)

    Menon, M.M.; Mioduszewski, P.K.; Owen, L.W.; Anderson, P.M.; Baxi, C.B.; Langhorn, A.; Luxon, J.L.; Mahdavi, M.A.; Schaffer, M.J.; Schaubel, K.M.; "" class="author-name" title=" (General Atomics Co., San Diego, CA (United States))" data-affiliation=" (General Atomics Co., San Diego, CA (United States))" >Smith, J.P>

    1992-01-01

    In this paper, a particle exhaust scheme using a cryocondensation pump in the advanced divertor configuration of the DIII-D tokamak is described. In this configuration, the pump is located inside a baffle chamber within the tokamak, designed to receive particles reflected off the divertor strike region. A concentric coaxial loop with forced-convection flow of two-phase helium is selected as the cryocondensation surface. The pumping configuration is optimized by Monte Carlo techniques to provide maximum exhaust efficiency while minimizing the deleterious effects of impingement of energetic plasma particles on cryogenic surfaces. Heat loading contributions from various sources on the cryogenic surfaces are estimated, based on which the cryogenic surfaces are estimated, based on which the cryogenic flow loop for the pump is designed. The mechanical aspects of the pump, designed to meet the many challenging requirements of operating the cryopump internal to the tokamak vacuum and in close proximity with the high-temperature plasma, are also outlined

  11. A DESIGN RETROSPECTIVE OF THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    LUXON, J.L

    2001-06-01

    OAK-B135 The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and rf heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research program. This paper gives an integrated picture of the facility and its capabilities

  12. NCSX Vacuum Vessel Fabrication

    International Nuclear Information System (INIS)

    Viola ME; Brown T; Heitzenroeder P; Malinowski F; Reiersen W; Sutton L; Goranson P; Nelson B; Cole M; Manuel M; McCorkle D.

    2005-01-01

    The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120 o vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1-inch of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120 o vessel segments are formed by welding two 60 o segments together. Each 60 o segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8-inch (20.3 cm) wide spacer ''spool pieces''. The vessel must have a total leak rate less than 5 X 10 -6 t-l/s, magnetic permeability less than 1.02(micro), and its contours must be within 0.188-inch (4.76 mm). It is scheduled for completion in January 2006

  13. PDX vacuum vessel stress analysis

    International Nuclear Information System (INIS)

    Nikodem, Z.D.

    1975-01-01

    A stress analysis of PDX vacuum vessel is described and the summary of results is presented. The vacuum vessel is treated as a toroidal shell of revolution subjected to an internal vacuum. The critical buckling pressure is calculated. The effects of the geometrical discontinuity at the juncture of toroidal shell head and cylindrical outside wall, and the concavity of the cylindrical wall are examined. An effect of the poloidal field coil supports and the vessel outside supports on the stress distribution in the vacuum vessel is determined. A method evaluating the influence of circular ports in the vessel wall on the stress level in the vessel is outlined

  14. The DIII-D cryogenic system upgrade

    International Nuclear Information System (INIS)

    Schaubel, K.M.; Laughon, G.J.; Campbell, G.L.; Langhorn, A.R.; Stevens, N.C.; Tupper, M.L.

    1993-10-01

    The original DIII-D cryogenic system was commissioned in 1981 and was used to cool the cryopanel arrays for three hydrogen neutral beam injectors. Since then, new demands for liquid helium have arisen including: a fourth neutral beam injector, ten superconducting magnets for the electron cyclotron heating gyrotrons, and more recently, the advanced diverter cryopump which resides inside the tokamak vacuum vessel. The original cryosystem could not meet these demands. Consequently, the cryosystem was upgraded in several phases to increase capacity, improve reliability, and reduce maintenance. The majority of the original system has been replaced with superior equipment. The capacity now exists to support present as well as future demands for liquid helium at DIII-D including a hydrogen pellet injector, which is being constructed by Oak Ridge National Laboratory. Upgrades to the cryosystem include: a recently commissioned 150 ell/hr helium liquefier, two 55 g/sec helium screw compressors, a fully automated 20-valve cryogen distribution box, a high efficiency helium wet expander, and the conversion of equipment from manual or pneumatic to programmable logic controller (PLC) control. The distribution box was designed and constructed for compactness due to limited space availability. Overall system efficiency was significantly improved by replacing the existing neutral beam reliquefier Joule-Thomson valve with a reciprocating wet expander. The implementation of a PLC-based automatic control system has resulted in increased efficiency and reliability. This paper will describe the cryosystem design with emphasis on newly added equipment. In addition, performance and operational experience will be discussed

  15. The DIII-D cryogenic system upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Schaubel, K.M.; Laughon, G.J.; Campbell, G.L.; Langhorn, A.R.; Stevens, N.C.; Tupper, M.L.

    1993-10-01

    The original DIII-D cryogenic system was commissioned in 1981 and was used to cool the cryopanel arrays for three hydrogen neutral beam injectors. Since then, new demands for liquid helium have arisen including: a fourth neutral beam injector, ten superconducting magnets for the electron cyclotron heating gyrotrons, and more recently, the advanced diverter cryopump which resides inside the tokamak vacuum vessel. The original cryosystem could not meet these demands. Consequently, the cryosystem was upgraded in several phases to increase capacity, improve reliability, and reduce maintenance. The majority of the original system has been replaced with superior equipment. The capacity now exists to support present as well as future demands for liquid helium at DIII-D including a hydrogen pellet injector, which is being constructed by Oak Ridge National Laboratory. Upgrades to the cryosystem include: a recently commissioned 150 {ell}/hr helium liquefier, two 55 g/sec helium screw compressors, a fully automated 20-valve cryogen distribution box, a high efficiency helium wet expander, and the conversion of equipment from manual or pneumatic to programmable logic controller (PLC) control. The distribution box was designed and constructed for compactness due to limited space availability. Overall system efficiency was significantly improved by replacing the existing neutral beam reliquefier Joule-Thomson valve with a reciprocating wet expander. The implementation of a PLC-based automatic control system has resulted in increased efficiency and reliability. This paper will describe the cryosystem design with emphasis on newly added equipment. In addition, performance and operational experience will be discussed.

  16. High temperature outgassing tests on materials used in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Holtrop, K.L.; Hansink, M.J.

    2006-01-01

    This article is a continuation of previous work on determining the outgassing characteristics of materials used in the DIII-D magnetic fusion tokamak [K. L. Holtrop, J. Vac. Sci. Technol. A 17, 2064 (1999)]. Achievement of high performance plasma discharges in the DIII-D tokamak requires careful control of impurity levels. Among the techniques used to control impurities are routine bakes of the vacuum vessel to an average temperature of 350 deg. C. Materials used in DIII-D must release only very small amounts of impurities (below 2x10 -6 mole) at this temperature that could be transferred to the first wall materials and later contaminate plasma discharges. To better study the behavior of materials proposed for use in DIII-D at elevated temperatures, the initial outgassing test chamber was improved to include an independent heating control of the sample and a simple load lock chamber. The goal was to determine not only the total degassing rate of the material during baking, but to also determine the gas species composition and to obtain a quantitative estimate of the degassing rate of each species by the use of a residual gas analyzer. Initial results for aluminum anodized using three different processes, stainless steel plated with black oxide and black chrome, and a commercially available fiber optic feedthrough will be presented

  17. Recent results from the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petersen, P.I.

    1998-02-01

    The DIII-D national fusion research program focuses on establishing the scientific basis for optimization of the tokamak approach to fusion energy production. The symbiotic development of research, theory, and hardware continues to fuel the success of the DIII-D program. During the last year, a radiative divertor and a second cryopump were installed in the DIII-D vacuum vessel, an array of central and boundary diagnostics were added, and more sophisticated computer models were developed. These new tools have led to substantial progress in the understanding of the plasma. The authors now have a better understanding of the divertor as a means to manage the heat, particle, and impurity transport pumping of the plasma edge using the in situ divertor cryopumps effectively controls the plasma density. The evolution of diagnostics that probe the interior of the plasma, particularly the motional Stark effect diagnostic, has led to a better understanding of the core of the plasma. This understanding, together with tools to control the profiles, including electron cyclotron waves, pellet injection, and neutral beam injection, has allowed them to progress in making plasma configurations that give rise to both low energy transport and improved stability. Most significant here is the use of transport barriers to improve ion confinement to neoclassical values. Commissioning of the first high power (890 kW) 110 GHz gyrotron validates an important tool for managing the plasma current profile, key to maintaining the transport barriers. An upgraded plasma control system, ''isoflux control,'' which exploits real time MHD equilibrium calculations to determine magnetic flux at specified locations within the tokamak vessel and provides the means for precisely controlling the plasma shape and, in conjunction with other heating and fueling systems, internal profiles

  18. Recent DIII-D results

    International Nuclear Information System (INIS)

    Petersen, P.I.

    1994-07-01

    This paper summarizes the recent DIII-D experimental results and the development of the relevant hardware systems. The DIII-D program focuses on divertor solutions for next generation tokamaks such as International Thermo-nuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX), and on developing configurations with enhanced confinement and stability properties that will lead to a more compact and economical fusion reactor. The DIII-D program carries out this research in an integrated fashion

  19. The ground-fault detection system for DIII-D

    International Nuclear Information System (INIS)

    Scoville, J.T.; Petersen, P.I.

    1987-10-01

    This paper presents a discussion of the ground-fault detection systems on the DIII-D tokamak. The subsystems that must be monitored for an inadvertent ground include the toroidal and poloidal coil systems, the vacuum vessel, and the coil support structures. In general, one point of each coil is tied to coil/power supply ground through a current limiting resistor. For ground protection the current through this resistor is monitored using a dynamically feedback balanced Hall probe transducer from LEM Industries. When large inductive currents flow in closed loops near the tokamak, the result is undesirable magnetic error fields in the plasma region and noise generation on signal cables. Therefore, attention must be paid to avoid closed loops in the design of the coil and vessel support structure. For DIII-D a concept of dual insulating breaks and a single-point ground for all structure elements was used to satisfy this requirement. The integrity of the support structure is monitored by a system which continuously attempts to couple a variable frequency waveform onto these single-point grounds. The presence of an additional ground completes the circuit resulting in current flow. A Rogowski coil is then used to track the unwanted ground path in order to eliminate it. Details of the ground fault detection circuitry, and a description of its operation will be presented. 2 refs., 7 figs

  20. Status of DIII-D plasma control

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Penaflor, B.

    1995-10-01

    A key component of the DIII-D Advanced Tokamak and Radiative Divertor Programs is the development and implementation of methods to actively control a large number of plasma parameters. These parameters include plasma shape and position, total stored energy, density, rf loading resistance, radiated power and more detailed control of the current profile. To support this research goal, a flexible and easily expanded digital control system has been developed and implemented. We have made parallel progress in modeling of the plasma, poloidal coils, vacuum vessel, and power system dynamics and in ensuring the integrity of diagnostic and command circuits used in control. Recent activity has focused on exploiting the mature digital control platform through the implementation of simple feedback controls of more exotic plasma parameters such as enhanced divertor radiation, neutral pressure and Marfe creation and more sophisticated identification and digital feedback control algorithms for plasma shape, vertical position, and safety factor on axis (q 0 ). A summary of recent progress in each of these areas will be presented

  1. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Hagiwara, Koji; Imura, Yasuya.

    1979-01-01

    Purpose: To provide constituted method for easily performing baking of vacuum vessel, using short-circuiting segments. Constitution: At the time of baking, one turn circuit is formed by the vacuum vessel and short-circuiting segments, and current transformer converting the one turn circuit into a secondary circuit by the primary coil and iron core is formed, and the vacuum vessel is Joule heated by an induction current from the primary coil. After completion of baking, the short-circuiting segments are removed. (Kamimura, M.)

  2. Currents in the DIII-D Tokamak

    Science.gov (United States)

    Azari, A.; Eidietis, N. W.

    2012-10-01

    Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

  3. Structural design of the DIII-D radiative divertor

    International Nuclear Information System (INIS)

    Reis, E.E.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Hollerbach, M.A.; Laughon, G.J.; Sevier, D.L.

    1996-10-01

    The divertor of the DIII-D tokamak is being modified to operate as a slot type, dissipative divertor. This modification, called the Radiative Divertor Program (RDP) is being carried out in two phases. The design and analysis is complete and hardware is being fabricated for the first phase. This first phase consists of an upper divertor baffle and cryopump to provide some density control for high triangularity, single or double null discharges. Installation of the first phase is scheduled to start in October, 1996. The second phase provides pumping at all four divertor strike points of double null high triangularity discharges and baffling of the neutral particles from transport back to the core plasma. Studies of the effects of varying the slot length and width of the divertor can be easily accomplished with the design of RDP hardware. Static and dynamic analyses of the baffle structures, new cryopumps, and feedlines were performed during the preliminary and final design phases. Disruption loads and differential thermal displacements must be accommodated in the design of these components. With the full RDP hardware installed, the plasma current in DIII-D will be a maximum of 3.0 MA. Plasma disruptions induce toroidal currents in the cryopump, producing complex dynamic loads. Simultaneously, the vacuum vessel vibrations impose a sinusoidal base excitation to the supports for the cryopump. Static and dynamic analyses of the cryopump demonstrate that the stresses due to disruption and thermal loadings satisfy the stress and deflection criteria

  4. Experimental survey of the L-H transition conditions in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Gohil, P.; Watkins, J.C.

    1994-01-01

    We present the global analysis of a recent survey of the H-mode power threshold in DIII-D using D o → D + NBI after boronization of the vacuum vessel. Single parameter scans of B T , I p , density, and plasma shape have been carried out on the DIII-D tokamak for neutral beam heated single-null and double-null diverted plasmas. In single-null discharges, the power threshold is found to increase approximately linearly with B T and n e but remains independent of I p . In double-null discharges, the power threshold is found to be approximately independent of both B T and n e . Various shape parameters such as plasma-wall gaps had only a weak effect on the power threshold. Imbalancing the double null configuration resulted in a large increase in the threshold power

  5. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Kurita, Gen-ichi; Onozuka, Masaki; Suzuki, Masaru.

    1997-01-01

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and γ rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  6. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Kurita, Gen-ichi [Japan Atomic Energy Research Inst., Tokyo (Japan); Onozuka, Masaki; Suzuki, Masaru

    1997-07-31

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and {gamma} rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  7. Wall stabilization of high beta plasmas in DIII-D

    International Nuclear Information System (INIS)

    Taylor, T.S.; Strait, E.J.; Lao, L.L.; Turnbull, A.D.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Groebner, R.J.; La Haye, R.J.; Mauel, M.

    1995-02-01

    Detailed analysis of recent high beta discharges in the DIII-D tokamak demonstrates that the resistive vacuum vessel can provide stabilization of low n magnetohydrodynamic (MHD) modes. The experimental beta values reaching up to β T = 12.6% are more than 30% larger than the maximum stable beta calculated with no wall stabilization. Plasma rotation is essential for stabilization. When the plasma rotation slows sufficiently, unstable modes with the characteristics of the predicted open-quotes resistive wallclose quotes mode are observed. Through slowing of the plasma rotation between the q = 2 and q = 3 surfaces with the application of a non-axisymmetric field, the authors have determined that the rotation at the outer rational surfaces is most important, and that the critical rotation frequency is of the order of Ω/2π = 1 kHz

  8. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Nagashima, Keisuke; Suzuki, Masaru; Onozuka, Masaki.

    1997-01-01

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  9. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Nagashima, Keisuke [Japan Atomic Energy Research Inst., Tokyo (Japan); Suzuki, Masaru; Onozuka, Masaki

    1997-07-11

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  10. DIII-D research operations

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D. (ed.)

    1993-05-01

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R D; and collaborative efforts.

  11. DIII-D research operations

    International Nuclear Information System (INIS)

    Baker, D.

    1993-05-01

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R ampersand D; and collaborative efforts

  12. DIII-D research operations

    International Nuclear Information System (INIS)

    La Haye, R.J.

    1994-05-01

    The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R ampersand D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma

  13. Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

    International Nuclear Information System (INIS)

    R.C. O'Neill; A.S. Bozek; M.E. Friend; C.B. Baxi; E.E. Reis; M.A. Mahdavi; D.G. Nilson; S.L. Allen; W.P. West

    1999-01-01

    The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo currents and toroidal induced currents. Incorporated also

  14. Vacuum vessel for plasma devices

    International Nuclear Information System (INIS)

    Yamada, Masao; Taguchi, Masami.

    1975-01-01

    Object: To permit effective utility of the space in the inner and outer sides of the container wall and also permit repeated assembly for use. Structure: Vacuum vessel wall sections are sealed together by means of welding bellows, and also flange portions formed at the end of the wall sections are coupled together by bolts and are sealed together with a seal ring and a seal cap secured by welding. (Nakamura, S.)

  15. RF power diagnostics and control on the DIII-D, 4 MW 30--120 MHz fast wave current drive system (FWCD)

    International Nuclear Information System (INIS)

    Ferguson, S.W.; Allen, J.C.; Callis, R.W.; Cary, W.P.; Harris, T.E.

    1995-10-01

    The Fast Wave Current Drive System uses three 2 MW transmitters to drive three antennas inside the DIII-D vacuum vessel. This paper describes the diagnostics for this system. The diagnostics associated with the General Atomics Fast Wave Current Drive System allow the system tuning to be analyzed and modified on a between shot basis. The transmitters can be exactly tuned to match the plasma with only one tuning shot into the plasma. This facilitates maximum rf power utilization

  16. DIII-D dust particulate characterization (June 1998 Vent)

    International Nuclear Information System (INIS)

    Carmack, W.J.

    1999-01-01

    Dust is a key component of fusion power device accident source term. Understanding the amount of dust expected in fusion power devices and its physical and chemical characteristics is needed to verify assumptions currently used in safety analyses. An important part of this safety research and development work is to characterize dust from existing experimental tokamaks. In this report, the authors present the collection, data analysis methods used, and the characterization of dust particulate collected from various locations inside the General Atomics DIII-D vacuum vessel following the June 1998 vent. The collected particulate was analyzed at the Idaho National Engineering and Environmental Laboratory (INEEL). Two methods were used to collect particulate with the goal of preserving the particle size distribution and physical characteristics of the particulate. Choice of collection technique is important because the sampling method used can bias the particle size distribution collected. Vacuum collection on substrates and adhesion removal with metallurgical replicating tape were chosen as non-intrusive sampling methods. Seventeen samples were collected including plasma facing surfaces in lower, upper, and horizontal locations, surfaces behind floor tiles, surfaces behind divert or tiles, and surfaces behind ceiling tiles. The results of the analysis are presented

  17. Progress of ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bayon, A. [F4E, c/ Josep Pla, No. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector 25, Gandhinagar 382025 (India); Preble, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

    2013-10-15

    Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

  18. Progress of ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Bayon, A.; Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B.; Kim, B.C.; Kuzmin, E.; Le Barbier, R.; Martinez, J.-M.; Pathak, H.; Preble, J.; Sa, J.W.; Terasawa, A.; Utin, Yu.

    2013-01-01

    Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure

  19. DIII-D power supply, design, and development

    International Nuclear Information System (INIS)

    Nerem, A.

    1995-02-01

    An overview of the DIII-D power supply system with information details concerning the configuration, power ratings, acquisition costs, and cost scaling relevant to the design of ITER and other tokamaks is presented. The power supplies for the DIII-D tokamak were installed and commissioned during the late 1970's and the beginning of the 1980's. Several upgrades have been implemented during the last two years to solve increasing reliability problems encountered as the equipment aged, to provide enhanced operational flexibilities, and to enable operation at the higher power levels needed to provide experimental data relevant to the ITER and TPX design activities. These upgrades ranged from redesign of the power supply control systems to the replacement of vacuum circuit breakers which had become unreliable in service. A new interlock and protection system has also been implemented using the latest programmable logic controllers (PLC) and computer technology. These upgrades have been highly successful and are described to provide insight to many issues in the specification of high power converters. Power supply models used in the design of the DIII-D Plasma Control System are also described along with model verification test data. These models are being used in the development of a new advanced plasma control system for the DIII-D tokamak. Recent operational experience and results are presented

  20. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

  1. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs

  2. Vertical stability, high elongation, and the consequences of loss of vertical control on DIII-D

    International Nuclear Information System (INIS)

    Kellman, A.G.; Ferron, J.R.; Jensen, T.H.; Lao, L.L.; Luxon, J.L.; Skinner, D.G.; Strait, E.J.; Reis, E.; Taylor, T.S.; Turnbull, A.D.; Lazarus, E.A.; Lister, J.B.

    1990-09-01

    Recent modifications to the vertical control system for DIII-D has enabled operation of discharges with vertical elongation κ, up to 2.5. When vertical stability is lost, a disruption follows and a large vertical force on the vacuum vessel is observed. The loss of plasma energy begins when the edge safety factor q is 2 but the current decay does not begin until q ∼1.3. Current flow on the open field lines in the plasma scrapeoff layer has been measured and the magnitude and distribution of these currents can explain the observed force on the vessel. Equilibrium calculations and simulation of this vertical displacement episode are presented. 7 refs., 4 figs

  3. Impact of environmental regulations on control of copper ion concentration in the DIII-D cooling water system

    International Nuclear Information System (INIS)

    Gootgeld, A.M.

    1993-10-01

    Tokamaks and industrial users are faced with the task of maintaining closed-loop, low conductivity, low impurity, cooling water systems. Operating these systems concentrates the impurities in the water requiring subsequent disposal. Environmental regulations are making this increasingly difficult. This paper will discuss the solution to the problem of removing and disposing of copper ions in the DIII-D low conductivity water system. Since the commissioning of the Doublet facility, the quality of the water in the 3000 gpm system that cools the DIII-D vacuum vessel coils, power supplies and auxiliary heating components has been controlled with mixed-bed ion exchangers. Low ion levels, particularly copper, are required to operate this equipment. In early 1992, the company that leases and regenerates DIII-D ion exchangers said they no longer can accept these resin beds for regeneration due to the level of copper ion on the resin. This change in policy, a change that has been adopted throughout their industry, was necessary to assure that the Metropolitan Sewerage System of the City of San Diego stays in compliance with State of California regulations and EPA-mandated national pretreatment standards and regulations. A cost effective solution was implemented which utilizes a reverse osmosis filtration system with the ion exchangers for make-up water. Levels of copper ion disposed to the sewer are in compliance with government standards. These measures have thus far proved effective in maintaining low conductivity and overall good quality cooling water. Specifically, this paper discusses DIII-D deionized cooling water quality requirements and an affective means to meet these requirements in order to be in compliance with government regulations for copper ion disposal. The problems discussed, the alternatives considered and the approach taken would be readily applicable to any deionized cooling water system containing copper where EPA standards and regulations are mandated

  4. A tangentially viewing VUV TV system for the DIII-D divertor

    International Nuclear Information System (INIS)

    Nilson, D.G.; Ellis, R.; Fenstermacher, M.E.; Brewis, G.; Jalufka, N.

    1998-07-01

    A video camera system capable of imaging VUV emission in the 120--160 nm wavelength range, from the entire divertor region in the DIII-D tokamak, was designed. The new system has a tangential view of the divertor similar to an existing tangential camera system which has produced two dimensional maps of visible line emission (400--800 nm) from deuterium and carbon in the divertor region. However, the overwhelming fraction of the power radiated by these elements is emitted by resonance transitions in the ultraviolet, namely the C IV line at 155.0 nm and Ly-α line at 121.6 nm. To image the ultraviolet light with an angular view including the inner wall and outer bias ring in DIII-D, a 6-element optical system (f/8.9) was designed using a combination of reflective and refractive optics. This system will provide a spatial resolution of 1.2 cm in the object plane. An intermediate UV image formed in a secondary vacuum is converted to the visible by means of a phosphor plate and detected with a conventional CID camera (30 ms framing rate). A single MgF 2 lens serves as the vacuum interface between the primary and secondary vacuums; a second lens must be inserted in the secondary vacuum to correct the focus at 155 nm. Using the same tomographic inversion method employed for the visible TV, they reconstruct the poloidal distribution of the UV divertor light. The grain size of the phosphor plate and the optical system aberrations limit the best focus spot size to 60 microm at the CID plane. The optical system is designed to withstand 350 C vessel bakeout, 2 T magnetic fields, and disruption-induced accelerations of the vessel

  5. Recycling and particle control in DIII-D

    International Nuclear Information System (INIS)

    Jackson, G.L.

    1991-11-01

    Particle control of both hydrogen and impurity atoms is important in obtaining reproducible discharges with a low fraction of radiated power in the DIII-D tokamak. The main DIII-D plasma facing components are graphite tiles and Inconel. Hydrogenic species desorbed from graphite during a tokamak discharge can be a major fueling source, especially in unconditioned graphite where these species can saturate the surface regions. In this case the recycling coefficient can exceed unity, leading to an uncontrolled density rise. In addition to removing volatile hydrocarbons and oxygen, DIII-D vessel conditioning efforts have been directed at the reduction of particle fueling from the graphite tiles. Conditioning techniques include: baking to ≤ 400 degrees C, low power pulsed discharge cleaning, and glow discharges in deuterium, helium, neon, or argon. Helium glow wall conditioning, is now routinely performed before every tokamak discharge. The effects of these techniques on hydrogen recycling and impurity influxes will be presented. The Inconel walls, while not generally exposed to high heat fluxes, nevertheless represent a source of metal impurities which can lead to impurity accumulation in the discharge and a high fraction of radiated power, particularly in H-mode discharges at higher plasma currents, I p > 1.5 MA. To reduce metal influx a thin (∼100 nm) low Z film has been applied on all plasma facing surfaces in DIII-D. The application of the boron film, referred to as boronization has the additional benefit over a carbon film of further reducing the oxygen influx. Following the first boronization in DIII-D a regime of very high confinement (VH-mode) was observed, characterized by low ohmic target density, low Z eff , and low radiated power

  6. Performance of the DIII-D neutral beam injection system

    International Nuclear Information System (INIS)

    Kim, J.; Callis, R.W.; Colleraine, A.P.; Cummings, J.; Glad, A.S.; Gootgeld, A.M.; Haskovec, J.S.; Hong, R.; Kellman, D.H.; Langhorn, A.R.

    1987-01-01

    During the upgrade of the Doublet III tokamak, the neutral beam injection system as also modified to accommodate long pulse sources and to utilize the larger entrance apertures to the torus vessel. All four beamlines on DIII-D are now in operation with a total of eight common long pulse sources. These have exhibited easier conditioning and good reproducibility. Performance results of the beamlines and supporting systems are presented, and the observed beam properties are discussed

  7. DIII-D DATA MANAGEMENT

    International Nuclear Information System (INIS)

    McHARG, B.B; BURUSS, J.R. Jr.; FREEMAN, J.; PARKER, C.T.; SCHACHTER, J.; SCHISSEL, D.P.

    2001-08-01

    OAK-B135 The DIII-D tokamak at the DIII-D National Fusion Facility routinely acquires ∼ 500 Megabytes of raw data per pulse of the experiment through a centralized data management system. It is expected that in FY01, nearly one Terabyte of data will be acquired. In addition there are several diagnostics, which are not part of the centralized system, which acquire hundreds of megabytes of raw data per pulse. There is also a growing suite of codes running between pulses that produce analyzed data, which add ∼ 10 Megabytes per pulse with total disk usage of about 100 Gigabytes. A relational database system has been introduced which further adds to the overall data load. In recent years there has been an order of magnitude increase in magnetic disk space devoted to raw data and a Hierarchical Storage Management system (HSM) was implemented to allow 7 x 24 unattended access to raw data. The management of all of the data is a significant and growing challenge as the quantities of both raw and analyzed data are expected to continue to increase in the future. This paper will examine the experiences of the approaches that have been taken in management of the data and plans for the continued growth of the data quantity

  8. Pumping Characteristics of the DIII-D Cryopump

    International Nuclear Information System (INIS)

    A.S. Bozek; C.B. Baxi; R.W. Callis; M.A. Mahdavi; R.C. O'Neill; E.E. Reis

    1999-01-01

    Beginning in 1992, the first of the DIII-D divertor baffles and cryocondensation pumps was installed. This open divertor configuration, located on the outermost floor of the DIII-D vessel, includes a cryopump with a predicted pumping speed of 50,000 ell/s excluding obstructions such as support hardware. Taking the pump structural and support characteristics into consideration, the corrected pumping speed for D 2 is 30,000 ell/s [1]. In 1996, the second divertor baffle and cryopump were installed. This closed divertor structure, located on the outermost ceiling of the DIII-D vessel, has a cryopump with a predicted pumping speed of 32,000 ell/s. In the fall of 1999, the third divertor baffle and cryopump will be installed. This divertor structure will be located on the 45 o angled corner on the innermost ceiling of the DIII-D vessel, known as the private flux region of the plasma configuration. With hardware supports factored into the pumping speed calculation, the private flux cryopump is expected to have a pumping speed of 15,000 ell/s. There was question regarding the effectiveness of the private flux cryopump due to the close proximity of the private flux baffle. This led to a conductance calculation study of the impact of rotating the cryopump aperture by 180 o to allow for greater particle and gas exhaust into the cryopump's helium panel. This study concluded that the cost and schedule impact of changing the private flux cryopump orientation and design did not warrant the possible 20% (3,000 ell/s) increase in pumping ability gained by rotating the cryopump aperture 180 o . The comparison of pumping speed of the first two cryocondensation pumps with the measured results will be presented as well as the calculation of the pumping speed for the private flux cryopump now being installed

  9. Baking results of KSTAR vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported.

  10. Baking results of KSTAR vacuum vessel

    International Nuclear Information System (INIS)

    Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M.

    2009-01-01

    The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported

  11. TORE SUPRA vacuum vessel and shield manufacturing

    International Nuclear Information System (INIS)

    Blateyron, J.; Lepez, R.

    1984-01-01

    TORE SUPRA vacuum vessel and vacuum chamber shield manufacturing in progress at Jeumont-Schneider consists of three main phases: - Detail engineering and manufacturing fixture construction; - Prototype section manufacturing and process preparation; - Construction of the 6 production modules. The welding techniques adopted, call for three special automatic processes: TIG, MIG and PLASMA welding which guarantee mechanical strength, vacuum tightness and absence of distortion. Production of the modules began July 1984. (author)

  12. Disruptions in DIII-D

    International Nuclear Information System (INIS)

    Reiman, A.; Taylor, P.; Kellman, A.; LaHaye, R.

    1996-01-01

    We report on the results of a statistical analysis of the DIII-D disruption data base, and on an examination of a selected subset of the shots to determine the likely causes of disruptions. The statistical analysis focuses on the dependence of the disruption rate on key dimensionless parameters. We find that the disruption frequency is high at modest values of the parameters, and that it can be relatively low at operational limits. For example, the disruption frequency in an ITER relevant regime (β N /l i ∼ 2, 3 G > 0.6, where n G is the Greenwald limit) is approximately 23%. For this range of q, the disruption frequency rises only modestly to about 35% at the β limit, consistent with previous observations of a soft β limit for this q regime. For the range 6 95 G G < .9) in all q regimes we have studied. The location of the minimum moves to higher density with increasing q

  13. Status and near-term plans for DIII-D

    International Nuclear Information System (INIS)

    Davis, L.G.; Callis, R.W.; Luxon, J.L.; Stambaugh, R.D.

    1987-10-01

    The DIII-D tokamak at GA Technologies began plasma operation in February of 1986 and is dedicated to the study of highly non-circular plasmas. High beta operation with enhanced energy confinement is paramount among the goals of the DIII-D research program. Commissioning of the device and facility has verified the design capability including coil and vessel loading, volt-second consumption, bakeout temperature, vessel armor, and neutral beamline thermal integrity and control systems performance. Initial experimental results demonstrate the DIII-D is capable of attaining high confinement (H-mode) discharges in a divertor configuration using modest neutral beam heating or ECH. Record values of I/sub p/aB/sub T/ have been achieved with ohmic heating as a first step toward operation at high values of toroidal beta and record values of beta have been achieved using neutral beam heating. This paper summarizes results to date and gives the near term plans for the facility. 13 refs., 6 figs., 1 tab

  14. Design of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

    1995-01-01

    The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

  15. Structural Analysis of the NCSX Vacuum Vessel

    International Nuclear Information System (INIS)

    Fred Dahlgren; Art Brooks; Paul Goranson; Mike Cole; Peter Titus

    2004-01-01

    The NCSX (National Compact Stellarator Experiment) vacuum vessel has a rather unique shape being very closely coupled topologically to the three-fold stellarator symmetry of the plasma it contains. This shape does not permit the use of the common forms of pressure vessel analysis and necessitates the reliance on finite element analysis. The current paper describes the NCSX vacuum vessel stress analysis including external pressure, thermal, and electro-magnetic loading from internal plasma disruptions and bakeout temperatures of up to 400 degrees centigrade. Buckling and dynamic loading conditions are also considered

  16. High Field Side Lower Hybrid Current Drive Launcher Design for DIII-D

    Science.gov (United States)

    Wallace, G. M.; Leccacori, R.; Doody, J.; Vieira, R.; Shiraiwa, S.; Wukitch, S. J.; Holcomb, C.; Pinsker, R. I.

    2017-10-01

    Efficient off-axis current drive scalable to reactors is a key enabling technology for a steady-state tokamak. Simulations of DIII-D discharges have identified high performance scenarios with excellent lower hybrid (LH) wave penetration, single pass absorption and high current drive efficiency. The strategy was to adapt known launching technology utilized in previous experiments on C-Mod (poloidal splitter) and Tore Supra (bi-junction) and remain within power density limits established in JET and Tore Supra. For a 2 MW source power antenna, the launcher consists of 32 toroidal apertures and 4 poloidal rows. The aperture is 60 mm x 5 mm with 1 mm septa and the peak n| | is 2.7+/-0.2 for 90□ phasing. Eight WR187 waveguides are routed from the R-1 port down under the lower cryopump, under the existing divertor, and up the central column with the long waveguide dimension along the vacuum vessel. Above the inner strike point region, each waveguide is twisted to orient the long dimension perpendicular to the vacuum vessel and splits into 4 toroidal apertures via bi-junctions. To protect the waveguide, the inner wall radius will need to increase by 2.5 cm. RF, disruption, and thermal analysis of the latest design will be presented. Work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using User Facility DIII-D, under Award Number DE-FC02-04ER54698 and by MIT PSFC cooperative agreement DE-SC0014264.

  17. Design and Analysis of the Cryopump for the DIII-D Upper Divertor

    International Nuclear Information System (INIS)

    Reis, E.E.; Baxi, C.B.; Bozek, A.S.

    1999-01-01

    A cryocondensation pump for the upper inboard divertor on DIII-D is to be installed in the vacuum vessel in the fall of 1999. The cryopump removes neutral gas particles from the divertor and prevents recycling to the plasma. This pump is designed for a pumping speed of 18,000 ell/s at 0.4 mTorr. The cryopump is toroidally continuous to minimize inductive voltages and avoid electrical breakdown during disruptions. The cryopump consists of a 25 mm Inconel tube cooled by liquid helium and is surrounded by nitrogen cooled shields. A segmented ambient temperature radiation/particle shield protects the nitrogen shields. The pump is subjected to a steady state heat load of less than 10 W due to conduction and radiation heat transfer. The helium tube will be subjected to Joule heating of less than 300 J due to induced current and a particle load of less than 12 W during plasma operation. The thermal design of the cryopump requires that it be cooled by 5 g/s liquid helium at an inlet pressure of 115 kPa and a temperature of 4.35 K. Thermal analysis and tests show that the helium tube can absorb a transient heat load of up to 100 W for 10 s and still pump deuterium at 6.3 K. Disruptions induce toroidal currents in the helium line and nitrogen shields. These currents cross the rapidly changing magnetic fields, applying complex dynamic loads on the cryopump. The forces on the pump are extrapolated from magnetic measurements from DIII-D plasma disruptions and scaled to a 3 MA disruption. The supports for the nitrogen shield consist of a racetrack design, which are stiff for reacting the disruption loads, but are radially flexible to allow differential thermal displacements with the vacuum vessel. Static and dynamic finite element analyses of the cryopump show that the stresses and displacements over a range of disruption and thermal loadings are acceptable

  18. Divertor extreme ultraviolet (EUV) survey spectroscopy in DIII-D

    Science.gov (United States)

    McLean, Adam; Allen, Steve; Ellis, Ron; Jarvinen, Aaro; Soukhanovskii, Vlad; Boivin, Rejean; Gonzales, Eduardo; Holmes, Ian; Kulchar, James; Leonard, Anthony; Williams, Bob; Taussig, Doug; Thomas, Dan; Marcy, Grant

    2017-10-01

    An extreme ultraviolet spectrograph measuring resonant emissions of D and C in the lower divertor has been added to DIII-D to help resolve an 2X discrepancy between bolometrically measured radiated power and that predicted by boundary codes for DIII-D, JET and ASDEX-U. With 290 and 450 gr/mm gratings, the DivSPRED spectrometer, an 0.3 m flat-field McPherson model 251, measures ground state transitions for D (the Lyman series) and C (e.g., C IV, 155 nm) which account for >75% of radiated power in the divertor. Combined with Thomson scattering and imaging in the DIII-D divertor, measurements of position, temperature and fractional power emission from plasma components are made and compared to UEDGE/SOLPS-ITER. Mechanical, optical, electrical, vacuum, and shielding aspects of DivSPRED are presented. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698 and DE-AC52-07NA27344, and by the LLNL Laboratory Directed R&D Program, project #17-ERD-020.

  19. VUV Spectroscopy in DIII-D Divertor

    International Nuclear Information System (INIS)

    Alkesh Punjabi; Nelson Jalufka

    2004-01-01

    The research carried out on this grant was motivated by the high power emission from the CIV doublet at 155 nm in the DIII-D divertor and to study the characteristics of the radiative divertor. The radiative divertor is designed to reduce the heat load to the target plates of the divertor by reducing the energy in the divertor plasma using upstream scrape-off-layer (SOL) radiation. In some cases, particularly in Partially Detached Divertor (PDD) operations, this emission accounts for more than 50% of the total radiation from the divertor. In PDD operation, produced by neutral gas injection, the particle flow to the target plate and the divertor temperature are significantly reduced. A father motivation was to study the CIV emission distribution in the lower, open divertor and the upper baffled divertor. Two Vacuum Ultra Violet Tangential viewing Television cameras (VUV TTV) were constructed and installed in the upper, baffled and the lower, open divertor. The images recorded by these cameras were then inverted to produce two-dimensional distributions of CIV in the poloidal plane. Results obtained in the project are summarized in this report

  20. RESEARCH PROGRESS AND HARDWARE SYSTEMS AT DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    PETERSEN,P.I; THE DIII-D TEAM

    2003-10-01

    OAK-B135 During the last two years significant progress has been made in the scientific understanding of DIII-D plasmas. Much of this progress has been enabled by the addition of new hardware systems. The electron cyclotron (EC) system has been upgraded from 3 MW to 6 MW, by adding three 1 MW gyrotrons with diamond windows and three steerable launchers (PPPL). The new gyrotrons have been tested to 1.0 MW for 5 s. The system has been used to control the 3/2 and 2/1 neoclassical tearing modes and to locally heat the plasma and thereby indirectly control the current density. Electron cyclotron current drive ECCD has been used to directly affect the current density. A Li-beam diagnostic has been brought on-line for measuring the edge current density using Zeeman splitting. A set of 12 coils (1-coils), consisting of six picture frame coils each above and below the midplane, with a capability of 7 kA for 10 s has been installed inside the DIII-D vessel. These coils, along with the existing six C-coils, are used to apply non-axisymmetric fields to the plasma for both exciting and controlling plasma instabilities. The DIII-D digital plasma control system is now used to not just control the shape and location of the plasma but also the electron temperature, density, the NTMs, RWMs, plasma beta and disruption mitigation. Plasma disruption experiments are extended to mitigation of real time detected disruptions on DIII-D.

  1. Device of supporting a vacuum plasma vessel

    International Nuclear Information System (INIS)

    Kanoi, Minoru; Hori, Yasuro.

    1980-01-01

    Purpose: To improve the earthquake-resistance of a vacuum plasma vessel by equalizing the natural vibrations of a vibrating system formed by supporting mechanisms of the respective sectors of the vessel. Constitution: The vacuum plasma vessel is constructed of bellows interposed among a plurality of thick sector-like rings and the rings, which are respectively supported by supporting mechanisms. Thus, the vibrating systems are divided into the rings interposed with the bellows, arms as the supporting mechanisms, and posts. The natural vibrations of these vibrating systems are equalized to each other by suitably adjusting the configurations and the sized of the arms and the posts or the weight or the like of the rings. Therefore, the respective rings become vibrated at the natural vibrations equal to each other so as to largely reduce the stresses produced at both ends of the bellows. Accordingly, it can remarkably improve the earthquake-resistance of the entire plasma vessel. (Sekiya, K.)

  2. Initial conditioning of the TFTR vacuum vessel

    International Nuclear Information System (INIS)

    Dylla, H.F.; Blanchard, W.R.; Krawchuk, R.B.; Hawryluk, R.J.; Owens, D.K.

    1984-01-01

    We report on the initial conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel prior to the initiation of first plasma discharges, and during subsequent operation with high power ohmically-heated plasmas. Following evacuation of the 86 m 3 vessel with the 10 4 1/s high vacuum pumping system, the vessel was conditioned by a 15 A dc glow discharge in H 2 at a pressure of 5 mTorr. Rapid-pulse discharge cleaning was used subsequently to preferentially condition the graphite plasma limiters. The effectiveness of the discharge cleaning was monitored by measuring the exhaust rates of the primary discharge products (CO/C 2 H 4 , CH 4 , and H 2 O). After 175 hours of glow discharge treatment, the equivalent of 50 monolayers of C and O was removed from the vessel, and the partial pressures of impurity gases were reduced to the range of 10 -9 -10 -10 Torr

  3. The strongest magnetic barrier in the DIII-D tokamak and comparison with the ASDEX UG

    Science.gov (United States)

    Ali, Halima; Punjabi, Alkesh

    2013-05-01

    Magnetic perturbations in tokamaks lead to the formation of magnetic islands, chaotic field lines, and the destruction of flux surfaces. Controlling or reducing transport along chaotic field lines is a key challenge in magnetically confined fusion plasmas. A local control method was proposed by Chandre et al. [Nucl. Fusion 46, 33-45 (2006)] to build barriers to magnetic field line diffusion by addition of a small second-order control term localized in the phase space to the field line Hamiltonian. Formation and existence of such magnetic barriers in Ohmically heated tokamaks (OHT), ASDEX UG and piecewise analytic DIII-D [Luxon, J.L.; Davis, L.E., Fusion Technol. 8, 441 (1985)] plasma equilibria was predicted by the authors [Ali, H.; Punjabi, A., Plasma Phys. Control. Fusion 49, 1565-1582 (2007)]. Very recently, this prediction for the DIII-D has been corroborated [Volpe, F.A., et al., Nucl. Fusion 52, 054017 (2012)] by field-line tracing calculations, using experimentally constrained Equilibrium Fit (EFIT) [Lao, et al., Nucl. Fusion 25, 1611 (1985)] DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. This second-order approach is applied to the DIII-D tokamak to build noble irrational magnetic barriers inside the chaos created by the locked resonant magnetic perturbations (RMPs) (m, n)=(3, 1)+(4, 1), with m and n the poloidal and toroidal mode numbers of the Fourier expansion of the magnetic perturbation with amplitude δ. A piecewise, analytic, accurate, axisymmetric generating function for the trajectories of magnetic field lines in the DIII-D is constructed in magnetic coordinates from the experimental EFIT Grad-Shafranov solver [Lao, L, et al., Fusion Sci. Technol. 48, 968 (2005)] for the shot 115,467 at 3000 ms in the DIII-D. A symplectic mathematical map is used to integrate field lines in the DIII-D. A numerical algorithm [Ali, H., et al., Radiat. Eff. Def. Solids Inc. Plasma Sc. Plasma Tech. 165, 83

  4. Leak detection on the DIII-D tokamak using helium entrainment techniques

    International Nuclear Information System (INIS)

    Brooks, N.H.; Baxi, C.; Anderson, P.

    1988-01-01

    The entrainment of helium in a viscous gas flow was utilized to compartmentalize, and then to pinpoint, a leak across the inner skin of the double-walled DIII-D vacuum vessel. Inaccessible from the outside, the leak connected the cooling channels in the wall interspace with the primary vacuum chamber. By entraining helium in the pressurized flow from the single-pass gas circulation system, well-defined portions of the wall were exposed to helium without disassembly of the poorly accessible cooling manifolds. Varying the helium injection point permitted the localization of the leak to a single 30 0 toroidal sector of the vessel. The exact location of the leak was found from inside the vessel by spraying helium on suspect regions, while sweeping the contents of the cooling channels to the foreline of a Varian Contraflow leak detector with a 0.1 Pa m 3 /s flow of nitrogen. Flow speed calculations were used to predict the response time to entrained helium of the actual leak detection setup

  5. Dust Studies in DIII-D and TEXTOR

    International Nuclear Information System (INIS)

    Rudakov, D.; Litnovsky, A.; West, W.; Yu, J.; Boedo, J.; Bray, B.; Brezinsek, S.; Brooks, N.; Fenstermacher, M.; Groth, M.; Hollmann, E.; Huber, A.; Hyatt, A.; Krasheninnikov, S.; Lasnier, C.; Moyer, R.; Pigarov, A.; Philipps, V.; Pospieszezyk, A.; Smirnov, R.; Sharpe, J.; Solomon, W.; Watkins, J.; Wong, C.

    2008-01-01

    Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Energetic plasma disruptions produce significant amounts of dust. However, dust production by disruptions alone is insufficient to account for the estimated in-vessel dust inventory in DIII-D. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by injecting micron-size dust in plasma discharges. In DIII-D, a sample holder filled with ∼30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure (∼0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. Individual dust particles are observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction, consistent with the drag force from the deuteron flow and in agreement with modeling by the 3D DustT code. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. Dust is launched either in the beginning of a discharge or at the initiation of NBI, preferentially in a direction perpendicular to the toroidal magnetic field. At the given configuration of the launch, the dust did not penetrate

  6. Expanding plasma jet in a vacuum vessel

    International Nuclear Information System (INIS)

    Chutov, Yu.I.; Kravchenko, A.Yu.; Yakovetskij, V.S.

    1998-01-01

    The paper deals with numerical calculations of parameters of a supersonic quasi-neutral argon plasma jet expanding into a cylindrical vacuum vessel and interacting with its inner surface. A modified method of large particles was used, the complex set of hydrodynamic equations being broken into simpler components, each of which describes a separate physical process. Spatial distributions of the main parameters of the argon plasma jet were simulated at various times after the jet entering the vacuum vessel, the parameters being the jet velocity field, the full plasma pressure, the electron temperature, the temperature of heavy particles, and the degree of ionization. The results show a significant effect of plasma jet interaction on the plasma parameters. The jet interaction with the vessel walls may result e.g. in excitation of shock waves and rotational plasma motions. (J.U.)

  7. 2XIIB vacuum vessel: a unique design

    International Nuclear Information System (INIS)

    Hibbs, S.M.; Calderon, M.O.

    1975-01-01

    The 2XIIB mirror confinement experiment makes unique demands on its vacuum system. The confinement coil set encloses a cavity whose surface is comprised of both simple and compound curves. Within this cavity and at the core of the machine is the operating vacuum which is on the order of 10 -9 Torr. The vacuum container fits inside the cavity, presenting an inside surface suitable for titanium getter pumping and a means of removing the heat load imposed by incandescent sublimator wires. In addition, the cavity is constructed of nonmagnetic and nonconducting materials (nonmetals) to avoid distortion of the pulsed confinement field. It is also isolated from mechanical shocks induced in the machine's main structure when the coils are pulsed. This paper describes the design, construction, and operation of the 2XIIB high-vacuum vessel that has been performing successfully since early 1974

  8. The TPX vacuum vessel and in-vessel components

    International Nuclear Information System (INIS)

    Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

    1994-01-01

    The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

  9. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-01-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. (In the past DIII-D operated with an open divertor.) The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix 'strike' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. (author) 5 refs., 5 figs

  10. Magnetic barriers and their q95 dependence at DIII-D

    Science.gov (United States)

    Volpe, F. A.; Kessler, J.; Ali, H.; Evans, T. E.; Punjabi, A.

    2012-05-01

    It is well known that externally generated resonant magnetic perturbations (RMPs) can form islands in the plasma edge. In turn, large overlapping islands generate stochastic fields, which are believed to play a role in the avoidance and suppression of edge localized modes (ELMs) at DIII-D. However, large coalescing islands can also generate, in the middle of these stochastic regions, KAM surfaces effectively acting as ‘barriers’ against field-line dispersion and, indirectly, particle diffusion. It was predicted in Ali and Punjabi (2007 Plasma Phys. Control. Fusion 49 1565-82) that such magnetic barriers can form in piecewise analytic DIII-D plasma equilibria. In this work, the formation of magnetic barriers at DIII-D is corroborated by field-line tracing calculations using experimentally constrained EFIT (Lao et al 1985 Nucl. Fusion 25 1611) DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. According to these calculations, the occurrence and location of magnetic barriers depend on the edge safety factor q95. It was thus suggested that magnetic barriers might contribute to narrowing the edge stochastic layer and play an indirect role in the RMPs failing to control ELMs for certain values of q95. The analysis of DIII-D discharges where q95 was varied, however, does not show anti-correlation between barrier formation and ELM suppression.

  11. Magnetic barriers and their q95 dependence at DIII-D

    International Nuclear Information System (INIS)

    Volpe, F.A.; Kessler, J.; Ali, H.; Punjabi, A.; Evans, T.E.

    2012-01-01

    It is well known that externally generated resonant magnetic perturbations (RMPs) can form islands in the plasma edge. In turn, large overlapping islands generate stochastic fields, which are believed to play a role in the avoidance and suppression of edge localized modes (ELMs) at DIII-D. However, large coalescing islands can also generate, in the middle of these stochastic regions, KAM surfaces effectively acting as ‘barriers’ against field-line dispersion and, indirectly, particle diffusion. It was predicted in Ali and Punjabi (2007 Plasma Phys. Control. Fusion 49 1565–82) that such magnetic barriers can form in piecewise analytic DIII-D plasma equilibria. In this work, the formation of magnetic barriers at DIII-D is corroborated by field-line tracing calculations using experimentally constrained EFIT (Lao et al 1985 Nucl. Fusion 25 1611) DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. According to these calculations, the occurrence and location of magnetic barriers depend on the edge safety factor q 95 . It was thus suggested that magnetic barriers might contribute to narrowing the edge stochastic layer and play an indirect role in the RMPs failing to control ELMs for certain values of q 95 . The analysis of DIII-D discharges where q 95 was varied, however, does not show anti-correlation between barrier formation and ELM suppression. (paper)

  12. Structural analysis of the KSTAR vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    In, Sang Ryul; Yoon, Byeong Joo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    Structure analysis of the vacuum vessel for the KSTAR tokamak which, is in the end phase of the conceptual design have been performed. Mechanical stresses and deformations of the vessel produced by constant forces due to atmospheric pressure, dead weight, fluid pressure, etc and various transient electromagnetic forces induced during tokamak operations were calculated as well as modal characteristics and buckling properties were investigated. Influences of the temperature gradient and the constraint condition of the support on the thermal stress and deformation of the vessel were analyzed. The thermal stress due to the temperature distribution on the vessel as supplying the N{sub 2} gas of 400 deg C through poloidal channels according to the recent baking concept were calculated. No severe problem in the robustness of the vessel was found when applying the constant pressures on the vessel. However the mechanical stress due to the EM force induced by halo currents flowing on the vessel and the plasma facing components (PFCs) far exceeded the allowable limit. Some reinforcing components should be added on the boundary of the PFC support and the vessel, and that of the vessel support and the vessel. A steep temperature gradient in the vicinity of the inlet and oulet of the heating gas produced a thermal stress much higher than allowable. It is necessary to make the temperature of the vessel as uniform as possible and to develop a new support concept which is flexible enough to accommodate a thermal expansion of a few cm while sufficiently strong to resist mechanical impacts. (author). 5 refs., 41 figs., 9 tabs.

  13. Protecting Against Damage from Refraction of High Power Microwaves in the DIII-D Tokamak

    Directory of Open Access Journals (Sweden)

    Lohr John

    2017-01-01

    Full Text Available Several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps have been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.

  14. Status of the ITER vacuum vessel construction

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C.H.; Sborchia, C.; Ioki, K.; Giraud, B.; Utin, Yu.; Sa, J.W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wang, X., E-mail: xiaoyuwww@gmail.com [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Teissier, P.; Martinez, J.M.; Le Barbier, R.; Jun, C.; Dani, S.; Barabash, V.; Vertongen, P.; Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Jucker, P.; Bayon, A. [F4E, c/ Josep Pla, n. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Pathak, H.; Raval, J. [ITER-India, IPR, A-29, Electronics Estate, GIDC, Sector-25, Gandhinagar 382025 (India); Ahn, H.J. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2014-10-15

    Highlights: • Final design of the ITER vacuum vessel (VV). • Procurement of the ITER VV. • Manufacturing results of real scale mock-ups. • Manufacturing status of the VV in domestic agencies. - Abstract: The ITER vacuum vessel (VV) is under manufacturing by four domestic agencies after completion of engineering designs that have been approved by the Agreed Notified Body (ANB). Manufacturing designs of the VV have been being completed, component by component, by accommodating requirements of the RCC-MR 2007 edition. Manufacturing of the VV first sector has been started in February 2012 in Korea and in-wall shielding in May 2013 in India. EU will start manufacturing of its first sector from September 2013 and Russia the upper port by the end of 2013. All DAs have manufactured several mock-ups including real-size ones to justify/qualify and establish manufacturing techniques and procedures.

  15. IMPROVEMENTS TO THE CRYOGENIC CONTROL SYSTEM ON DIII-D

    International Nuclear Information System (INIS)

    HOLTROP, K.L; ANDERSON, P.M; MAUZEY, P.S.

    2004-03-01

    OAK-B135 The cryogenic facility that is part of the DIII-D tokamak system supplies liquid nitrogen and liquid helium to the superconducting magnets used for electron cyclotron heating, the D 2 pellet injection system, cryopumps in the DIII-D vessel, and cryopanels in the neutral beam injection system. The liquid helium is liquefied on site using a Sulzer liquefier that has a 150 l/h liquefaction rate. Control of the cryogenic facility at DIII-D was initially accomplished through the use of three different programmable logic controllers (PLCs). Recently, two of those three PLCs, a Sattcon PLC controlling the Sulzer liquefier and a Westinghouse PLC, were removed and all their control logic was merged into the remaining PLC, a Siemens T1555. This replacement was originally undertaken because the removed PLCs were obsolete and unsupported. However, there have been additional benefits from the replacement. The replacement of the RS-232 serial links between the graphical user interface and the PLCs with a high speed Ethernet link allows for real-time display and historical trending of nearly all the cryosystem's data. this has greatly increased the ability to troubleshoot problems with the system, and has permitted optimization of the cryogenic system's performance because of the increased system integration. To move the control logic of the Sattcon control loops into the T1555, an extensive modification of the basic PID control was required. These modifications allow for better control of the control loops and are now being incorporated in other control loops in the system

  16. Development, installation, and initial operation of DIII-D graphite armor tiles

    International Nuclear Information System (INIS)

    Anderson, P.M.; Baxi, C.B.; Reis, E.E.; Smith, J.P.; Smith, P.D.

    1988-04-01

    An upgrade of the DIII-D vacuum vessel protection system has been completed. The ceiling, floor, and inner wall have been armored to enable operation of CIT-relevant doublenull diverted plasmas and to enable the use of the inner wall as a limiting surface. The all- graphite tiles replace the earlier partial coverage armor configuration which consisted of a combination of Inconel tiles and graphite brazed to Inconel tiles. A new all-graphite design concept was chosen for cost and reliability reasons. The 10 minute duration between plasma discharges required the tiles to be cooled by conduction to the water-cooled vessel wall. Using two and three- dimensional analyses, the tile design was optimized to minimize thermal stresses with uniform thermal loading on the plasma-facing surface. Minimizing the stresses around the tile hold-down feature and eliminating stress concentrators were emphasized in the design. The design of the tile fastener system resulted in sufficient hold-down forces for good thermal conductance to the vessel and for securing the tile against eddy current forces. The tiles are made of graphite, and a program to select a suitable grade of graphite was undertaken. Initially, graphites were compared based on published technical data. Graphite samples were then tested for thermal shock capacity in an electron beam test facility at the Sandia National Laboratory (SNLA) in Albuquerque, New Mexico, USA. 4 refs., 6 figs

  17. A fast scanning probe for DIII--D

    International Nuclear Information System (INIS)

    Watkins, J.G.; Salmonson, J.; Moyer, R.; Doerner, R.; Lehmer, R.; Schmitz, L.; Hill, D.N.

    1992-01-01

    A fast reciprocating probe has been developed for DIII--D which can penetrate the separatrix during H mode with up to 5 MW of NBI heating. The probe has been designed to carry various sensor tips into the scrape-off layer at a velocity of 3 m/s and dwell motionless for a programmed period of time. The driving force is provided by a pneumatic cylinder charged with helium to facilitate greater mass flow. The first series of experiments have been done using a Langmuir probe head with five graphite tips to measure radial profiles of n e , T e , φ f , n e , and φ f . The amplitude and phase of the fluctuating quantities are measured by using specially constructed vacuum compatible 5-kV coaxial transmission lines which allow us to extend the measurements into the MHz range. TTZ ceramic bearings and fast stroke bellows were also specially designed for the DIII--D probe. Initial measurements will be presented

  18. Installation and initial operation of the DIII-D advanced divertor cryocondensation pump

    International Nuclear Information System (INIS)

    Smith, J.P.; Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Hyatt, A.W.; Laughon, G.J.; Mahdavi, M.A.; Reis, E.E.; Schaffer, M.J.; Sevier, D.L.; Stambaugh, R.D.; Menon, M.M.

    1993-10-01

    Phase two of a divertor cryocondensation pump, the Advanced Divertor Program, is now installed in the DIII-D tokamak at General Atomics and complements the phase one biasable ring electrode. The installation consists of a 10 m long cryocondensation pump located in the divertor baffle chamber to study plasma density control by pumping of the divertor. The design is a toroidally electrically continuous liquid helium-cooled panel with 1 m 2 of pumping surface. The helium panel is single point grounded to the nitrogen shield to minimize eddy currents. The nitrogen shield is toroidally continuous and grounded to the vacuum vessel in 24 locations to prevent voltage potentials from building up between the pump and vacuum vessel wall. A radiation/particle shield surrounds the nitrogen-cooled surface to minimize the heat load and prevent water molecules condensed on the nitrogen surface from being released by impact of energetic particles. Large currents (>5000 A) are driven in the helium and nitrogen panels during ohmic coil ramp up and during disruptions. The pump is designed to accommodate both the thermal and mechanical loads due to these currents. A feedthrough for the cryogens allows for both radial and vertical motion of the pump with respect to the vacuum vessel. Thermal performance measured on a prototype verified the analytical model and thermal design of the pump. Characterization tests of the installed pump show the pumping speed in deuterium is 42,000 ell/sec for a pressure of 5 mTorr. Induction heating of the pump (at 300 W) resulted in no degradation of pumping speed. Plasma operations with the cryopump show a 60% lower density in H-mode

  19. Vacuum vessel for a nuclear fusion device

    International Nuclear Information System (INIS)

    Watanabe, Takashi; Sato, Hiroshi; Owada, Koro.

    1976-01-01

    Object: To provide a reinforcing member on a bellows portion to reduce a stress at the bellows portion thereby increasing the strength of a vessel. Structure: A vacuum vessel for a nuclear fusion device has a bellows portion and a wall thick portion. A support extended toward the bellows portion is secured inside of a toroidal section in order to reduce the stress at the bellows portion. An insulator is interposed between the support and the bellows portion and is retained on the support by a bolt. Since the stress may be reduced by the support, the wall thick of the bellows portion may be decreased to sufficiently secure the low electric resistance value. (Yoshihara, H.)

  20. 110GHz ECH on DIII-D

    International Nuclear Information System (INIS)

    Cary, W.P.; Allen, J.C.; Callis, R.W.; Doane, J.L.; Harris, T.E.; Moetler, C.P.; Neren, A.; Prater, P.; Rensen, D.

    1992-01-01

    This paper reports on a new high power electron cyclotron heating (ECH) system which 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 second 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 the 30-60 MHz ICH (2MW) systems. H-mode physics, plasma stabilization experiments and transport studies are also to be conducted at 110 GHz

  1. DIII-D physics analysis database

    International Nuclear Information System (INIS)

    Bramson, G.; Schissel, D.P.; DeBoo, J.C.; St John, H.

    1990-10-01

    Since June 1986 the DIII-D tokamak has had over 16000 discharges accumulating more than 250 Gigabytes of raw data (currently over 30 Mbytes per discharge). The centralized DIII-D databases and the associated support software described earlier provide the means to extract, analyze, store, and display reduced sets of data for specific physics issues. The confinement, stability, transition, and cleanliness databases consist of more than 7500 records of basic reduced diagnostic data datasets. Each database record corresponds to a specific snapshot in time for a selected discharge. Recently some profile datasets have been implemented. Diagnostic data are fit by a cubic spline or a parabola by the in-house ENERGY code to provide density, temperature, radiated power, effective charge (Z eff ), and rotation velocity profiles. These fits are stored in the profile datasets which are inputs for the ONETWO code which computes transport data. 3 refs., 4 figs

  2. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-04-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix ''strike'' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. 5 refs., 5 figs

  3. Advanced tokamak physics in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)

    2000-12-01

    Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)

  4. Disruption studies in DIII-D

    International Nuclear Information System (INIS)

    Kellman, A.G.; Evans, T.E.; Cuthbertson, J.W.

    1996-09-01

    Characteristics of disruptions in the DIII-D tokamak including the current decay rate, halo current magnitude and toroidal asymmetry, and heat pulse to the divertor are described. Neon and argon pellet injection is shown to be an effective method for mitigating the halo currents and the heat pulse with a 50% reduction in both quantities achieved. The injection of these impurity pellets frequently gives rise to runaway electrons

  5. Shielding performance of the NET vacuum vessel

    International Nuclear Information System (INIS)

    Arkuszewski, J.J.; Jaeger, J.F.

    1988-01-01

    To corroborate 1-D deterministic shielding calculations on the Next European Torus (NET) vacuum vessel/shield and shielding blanket, 3-D Monte Carlo calculations have been done with the MCNP code. This should provide information on the poloidal and the toroidal variations. Plasma source simulation and the geometrical model are described, as are other assumptions. The calculations are based on the extended plasma power of 714 MW. The results reported here are the heat deposition in various parts of the device, on the one hand, and the neutron and photon currents at the outer boundary of the vacuum vessel, on the other hand. The latter are needed for the detailed design of the super-conducting magnetic coils. A reasonable statistics has been obtained on the outboard side of the torus, though this cannot be said for the inboard side. The inboard is, however, much more toroidally symmetric than the outboard, so that other methods could be applied such as 2-D deterministic calculations, for instance. (author) 4 refs., 44 figs., 42 tabs

  6. DIII-D edge physics database

    International Nuclear Information System (INIS)

    Jong, R.A.; Porter, G.D.; Hill, D.N.; Buchenauer, D.A.; Bramson, G.

    1992-03-01

    We have developed an edge-physics database containing data for the plasma in the divertor region and the scrape-off layer (SOL) for the DIII-D tokamak. The database provides many of the parameters necessary to model the power flow to the divertor and other plasma processes in the plasma edge. It will also facilitate the analysis of DIII-D data for comparison with other divertor tokamaks. In addition to the core plasma parameters, edge-specific data are included in this database. Initial results using the database show good agreement between the pressure profiles measured by the Langmuir probes and those determined from the Thomson data for the inner strike point, but not for the outer strike point region. We also find that the ratio of separatrix density to average core density, as well as the in/out asymmetry in the SOL power at the divertor in DIII-D do not agree with values currently assumed in modeling the International Thermonuclear Experimental Reactor (ITER)

  7. Motional stark effect upgrades on DIII-D

    International Nuclear Information System (INIS)

    Rice, B.W.; Nilson, D.G.; Wroblewski, D.

    1994-04-01

    The measurement and control of the plasma current density profile (or q profile) is critical to the advanced tokamak program on DIII-D. A complete understanding of the stability and transport properties of advanced operating regimes requires detail poloidal field measurements over the entire plasma radius from the core to the edge. In support of this effort, the authors have recently completed an upgrade of the existing MSE diagnostic, increasing the number of channels from 8 to 16. A new viewing geometry has been added to the outer edge of the plasma which improves the radial resolution in this region from 10 cm to < 4 cm. This view requires the use of a reflector that has been designed to minimize polarization amplitude and phase effects. Vacuum-compatible polarizers have also been added to the instrument for in-situ calibration. Future use of the MSE diagnostic for feedback control of the q profile will also be discussed

  8. A cryocondensation pump for the DIII-D Advanced Divertor Program

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.; Reis, E.; Sevier, L.

    1992-03-01

    A cryocondensation pump was designed for the baffle chamber of General Atomics DIII-D tokamak and will be installed in the fall of 1992. The purpose of the pump is to study plasma density control by pumping the divertor. The pump is toroidally continuous, approximately 10 m long and located in the lower outer corner of the vacuum chamber of the machine. It consists of a 1 m 2 liquid helium-cooled surface surrounded by a liquid nitrogen-cooled shield to limit the heat load on the helium-cooled surface. The liquid nitrogen-cooled surface is surrounded by a radiation/particle shield to prevent energetic particles from impacting and releasing condensed water molecules. A thermal enhancement coating was applied to the nitrogen shell to lower the maximum temperature of the shell. The coating is non-continuous to keep the toroidal electrical resistance high. The whole pump is supported off the water-cooled vacuum vessel wall. Supports for the pump were designed to accommodate the thermal differences between the 4 K helium surface, the 77 K nitrogen shells, and the 300 K vacuum vessel supporting the pump and to provide a low heat leak structural support. Disruption loading on the pump was analyzed and a finite element structural analysis of the pump was completed. A testing program was completed to evaluate coating techniques to enhance heat transfer and emissivity of the various surfaces. Fabrication tests were performed to determine the best method of attaching the liquid nitrogen flow tubes to their shield surfaces and to determine the best alternative to fabricating the different shells of the pump. A prototype sector of the pump was built to verify fabrication and assembly techniques

  9. Control of the Resistive Wall Mode with Internal Coils in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Okabayashi, M.; Bialek, J.; Bondeson, A.

    2005-01-01

    New coils were installed inside the vacuum vessel of the DIII-D device for producing nonaxisymmetric magnetic fields. These 'Internal-Coils' are predicted to stabilize the Resistive Wall Mode (RWM) branch of the long-wavelength external kink mode with plasma beta close to the ideal wall limit. Feedback using these new Internal-Coils was found to be more effective when compared with using the External-Coils located outside the vacuum vessel, because the location inside the vessel allows faster response and their geometry also couples better to the helical mode structure. A proper choice of feedback gain increased the plasma beta above the no-wall limit to C β ≥ 0.9, where C β is a measure of achievable beta above no-wall limit defined as (β-β no-wall.limit )/(β ideal.wall.limit )-)/(β no.wall.limit ). The feedback system with Internal-Coils can suppress the RWM up to the normalized growth rate γτ w > or ∼ 10 (τ w is the resistive flux penetration time of the wall). The feedback-driven dynamic error field correction helps to stabilize the RWM by reducing the rotational drag for Ω rot > Ω crit , where Ω rot is the angular rotation frequency of plasma and Ω crit is the critical value for the rotational stabilization. When Ω rot crit /2, the feedback system must stabilize the RWM mainly through direct magnetic control of the mode. The estimated Ω crit /Ω A is ∼ 2.5% by the MARS-F code analysis with experimentally observed profiles, where /Ω A is the Alfven angular rotational frequency at q 2 surface. The MARS-F code also predicts that for successful RWM magnetic feedback control the power supply characteristic time should be a fraction of the growth time of the targeted RWM. (author)

  10. Fabrication of the new poloidal field coils for DIII-D

    International Nuclear Information System (INIS)

    Heiberger, M.; Bott, R.J.; Gallix, R.; Street, R.W.

    1986-01-01

    The six new poloidal field coil assemblies manufactured by GA Technologies (GA) for DIII-D range in diameter from 3.4-5.3 m. Two of them are 55-turn field shaping coils. Each of the other four combines one turn of the ohmic heating coil and a 55-turn field shaping coil into a single unit encased in a stainless steel box beam. These four box beams, which provide support for the coils inside, are part of the overall coil and vacuum vessel support structure. They also serve as molds for vacuum impregnating the coils with epoxy. All coils are made of hollow, water-cooled copper conductor. The larger field shaping coils are designed for 20 kA, 3 sec rectangular current pulses with 40 0 C temperature rise. The ohmic heating coil turns are capable of currents of up to 110 kA. The conductor is wrapped with Kapton and fiberglass tape; Kapton provides 1000 V/turn and 28 kV coil-to-ground insulation. The fiberglass acts as wick and reinforcement for the vacuum impregnated epoxy resin which bonds the coil together. The fabrication process is described in detail and illustrated. Tools and setups used for special operations such as induction brazing, conductor winding, conductor bending, and vacuum impregnation are presented. The quality control procedures followed to guarantee sound brazed joints are explained. The electrical tests performed at several stages of fabrication, especially the 1000 V/turn impulse tests conducted before potting to facilitate fault detection and repair, are described

  11. Structural analysis of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sannazzaro, G.; Ioki, K.; Johnson, G.; Onozuka, M.; Utin, Y. [ITER Joint Work Site, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States); Swanson, J. [USHT, Raytheon, Princeton (United States)

    1998-07-01

    The ITER Vacuum Vessel (VV) must withstand a large number of loading conditions including electromagnetic, seismic, operational and upset pressure, thermal and test loads. All of the loading conditions and load combinations have been categorized and classified to permit the allowable stress to be defined in accordance with the recommendations of the ASME code. The most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and the load combination of seismic and electromagnetic loads due to a plasma vertical instability. The areas of high stress are the regions around the VV and the blanket supports, and the attachment of the ports to the main shell. In all of the loading conditions and load combinations the calculated stresses are below the allowable values. (authors)

  12. TPX vacuum vessel transient thermal and stress conditions

    International Nuclear Information System (INIS)

    Feldshteyn, Y.; Dinkevich, S.; Feng, T.; Majumder, D.

    1995-01-01

    The TPX vacuum vessel provides the vacuum boundary for the plasma and the mechanical support for the internal components. Another function of the vacuum vessel is to contain neutron shielding water in the double wall space during normal operation. This double wall space serves as a heat reservoir for the entire vacuum vessel during bakeout. The vacuum vessel and the internal components are subjected to thermal stresses induced by a nonuniform temperature distribution within the structure during bakeout. A successful Conceptual Design Review in March 1993 has established superheated steam as the heating source of the vacuum vessel. A transient bakeout mode of the vacuum vessel and in-vessel components has been analyzed to evaluate transient period duration, proper temperature level, actual thermal stresses and performance of the steam equipment. Thermally, the vacuum vessel structure may be considered as an adiabatic system because it is perfectly insulated by the strong surrounding vacuum and multiple layers of superinsulation. Important aspects of the analysis are described herein

  13. Thermal design, analysis, and experimental verification for a DIII-D cryogenic pump

    International Nuclear Information System (INIS)

    Baxi, C.B.; Anderson, P.; Langhorn, A.; Schaubel, K.; Smith, J.

    1991-01-01

    As part of the advanced divertor program, it is planned to install a 50 m 3 /s capacity cryopump for particle removal in the DIII-D tokamak. The cryopump will be located in the outer bottom corner of the vacuum vessel. The pump will consist of a surface at liquid helium temperature (helium panel) with a surface area of about 1 m 2 , a surface at liquid nitrogen temperature (nitrogen shield) to reduce radiation heat load on the helium panel, and a secondary shield around the nitrogen shield. The cryopump design poses a number of thermal hydraulic problems such as estimation of heat loads on helium and nitrogen panels, stability of the two-phase helium flow, performance of the pump components during high temperature bakeout, and cooldown performance of the helium panel from ambient temperatures. This paper presents the thermal analysis done to resolve these issues. A prototypic experiment performed at General Atomics verified the analysis and increased the confidence in the design. The experimental results are also summarized in this paper. (orig.)

  14. OEDGE modeling of the DIII-D double null (CH4)-C-13 puffing experiment

    International Nuclear Information System (INIS)

    Elder, J.D.; Wampler, W.R.; McLean, A.G.; Stangeby, P.C.; Allen, S.L.; Bray, B.D.; Brooks, N.H.; Leonard, A.W.; Unterberg, Ezekial A.; Watkins, J.G.

    2011-01-01

    Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. (CH4)-C-13 was injected into the secondary outer divertor to simulate sputtering there. The majority of the C-13 found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall (CH4)-C-13 injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 degrees C, is incomplete. The principal features of the measured C-13 deposition pattern have been replicated by the OEDGE interpretive code.

  15. The long range DIII-D plan

    International Nuclear Information System (INIS)

    Simonen, T.C.

    1993-02-01

    The mission of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. The National Energy Strategy calls for the development of magnetic fusion as an energy option with operation of a DEMO by 2025. The DEMO will be based on nuclear technology demonstrated in ITER and the physics and engineering database established in magnetic fusion facilities during the next two decades. On the present path, based on extrapolation of current conventional operating modes, ITER is twice as large as Joint European Tokamak (JET), and DEMO, using the same logic, will be even larger. However, successful development of advanced tokamak operating modes could open the way for significantly improved confinement and stability, leading to a smaller, more commercially attractive DEMO, provided new diverter concepts are developed to handle the accompanying high divertor power density. A smaller and lower cost DEMO opens up the possibility that multiple nations, utilities, and industries could build DEMOs simultaneously and, therefore, more rapidly optimize the tokamak for commercialization. Results from experiments at DIII-D and other tokamaks indicate that plasma and divertor performance can be increased transiently beyond the baseline conceptual design of ITER. A simultaneous long pulse demonstration of such improved tokamak plasma and divertor operation for steady state would establish an advanced physics foundation for the tokamak physics experiment program, provide new operating options for ITER, and open a path to an attractive DEMO. The planned DIII-D program incorporates new theory and technology developments to extend the tokamak experimental physics database toward steady state. This research program will also continue to provide increased understanding in many areas of fusion science and technology

  16. ICRH coupling in DIII-D

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Baity, F.W.; Bryan, W.E.; Jaeger, E.F.; Owens, T.L.; Remsen, D.B.; Luxon, J.; Rawls, J.M.

    1986-01-01

    A 9-MW ion cyclotron resonant frequency (ICRF) experiment has been proposed to heat the Doublet III-D (DIII-D) plasma. DIII-D is a 2.2-T, 3.5-MA tokamak at GA Technologies with a major radius of 1.67 m and minor radius of 67 cm (elongation approx.2). The device was recommissioned in early 1986. The initial experimental program includes ohmic plasma and neutral beam studies; high-power rf experiments will follow in later years. Compact loop antennas (which fit completely in a 35- by 50-cm port) have been chosen to convey this power because of their inherent ease of maintenance, high efficiency, and versatility. In order to verify that the antenna will have sufficient loading, a prototype low-power (2-MW) antenna has been designed and installed. Measurements will be made through September 1986. The antenna is a cavity antenna that will operate from approximately 30 to 80 MHz with a 50-Ω match for a load resistance of approx.1 Ω. It is surrounded by a fixed graphite-covered frame and can be extended from 3 cm behind this frame to 2 cm in front. This can be used to adjust coupling to the plasma. The electrical, mechanical, and thermal characteristics of this antenna system (and its extrapolation to ignited tokamaks) are discussed. In addition to experimental exploration of coupling, we have investigated wave propagation and absorption in DIII-D by using a cold collisional plasma model in straight tokamak geometry with rotation transform. Loading and power deposition profiles as a function of frequency, density, and species mix are presented

  17. DIII-D Advanced Tokamak Research Overview

    International Nuclear Information System (INIS)

    V.S. Chan; C.M. Greenfield; L.L. Lao; T.C. Luce; C.C. Petty; G.M. Staebler

    1999-01-01

    This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously β N H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues

  18. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Nakahira, Masataka

    2006-01-01

    The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

  19. Fabrication of the vacuum vessel for JT-60 machine upgrade

    International Nuclear Information System (INIS)

    Uchikawa, T.; Takanabe, K.; Tsujimura, S.; Ue, K.; Oka, K.; Kuri, S.; Ioki, K.; Namiki, K.; Suzuki, Y.; Horliike, H.; Ninomiya, H.; Yamamoto, M.; Neyatani, Y.; Ando, T.; Matsukawa, M.

    1992-01-01

    The JT-60 tokamak was upgraded to double the plasma current to 6 MA. In the JT-60 machine upgrade (JT-60U), the vacuum vessel and poloidal field (PF) coils were renewed. The new vacuum vessel features a three-dimensionally curved, thin double-skin torus with multi-arc D-shaped cross section. The double-skin structure is strengthened with square pipes placed in between the outer and inner skins. Fabrication and site installation of the vessel was smoothly completed in February, 1991. This paper describes an overview of the JT-60U vacuum vessel construction

  20. Commissioning of the long-pulse fast wave current drive antennas for DIII-D

    International Nuclear Information System (INIS)

    Baity, F.W.; Barber, G.C.; Goulding, R.H.; Hoffman, D.J.; DeGrassie, J.S.; Pinsker, R.I.; Petty, C.C.; Cary, W.

    1995-01-01

    Two new four-element fast wave current drive antennas have been installed on DIII-D. These antennas are designed for 10-s pulses at 2 MW each in the frequency range of 30 to 120 MHz. Each element comprises two poloidal segments fed in parallel in order to optimize plasma coupling at the upper end of the frequency range. The antennas are mounted on opposite sides of the vacuum vessel, in ports designated 0 degrees and 180 degrees after their toroidal angle. Each antenna array is fed by a single transmitter. The power is first split two ways by means of a 3-dB hybrid coupler, then each of these lines feeds a resonant loop connecting a pair of array elements. The power transfer during asymmetric phasing is shunted between resonant loops by a decoupler. The resonant loops are fitted with line stretchers so that multiple frequencies of operation are possible without reconfiguring the transmission line. Commissioning of these antennas has been underway since June 1994. Several deficiencies in the transmission line system were uncovered during initial vacuum conditioning, including problems with the transmission line insulators and with the drive rods for the variable elements. The former was solved by replacing the original alumina insulators, and the latter has been avoided during operation to date by positioning the tuners to avoid high voltage appearing on the drive rods. A modified design for the drive rods will be implemented before RF operations resume operation June 1995. New transmitters were procured from ABB for the new antennas and were installed in parallel with the antenna installation. During initial vacuum conditioning of the antenna in the 180 degree port a fast digital oscilloscope was used to try to pinpoint the location of arcing by a time-of-flight technique and to develop an understanding of the typical arc signature in the system

  1. Fast wave current drive on DIII-D

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I.

    1995-01-01

    The physics of electron heating and current drive with the fast magnetosonic wave has been demonstrated on DIII-D, in reasonable agreement with theoretical modeling. A recently completed upgrade to the fast wave capability should allow full noninductive current drive in steady state advanced confinement discharges and provide some current density profile control for the Advanced Tokamak Program. DIII-D now has three four-strap fast wave antennas and three transmitters, each with nominally 2 MW of generator power. Extensive experiments have been conducted with the first system, at 60 MHz, while the two newer systems have come into operation within the past year. The newer systems are configured for 60 to 120 MHz. The measured FWCD efficiency is found to increase linearly with electron temperature as γ = 0.4 x 10 18 T eo (keV) [A/m 2 W], measured up to central electron temperature over 5 keV. A newly developed technique for determining the internal noninductive current density profile gives efficiencies in agreement with this scaling and profiles consistent with theoretical predictions. Full noninductive current drive at 170 kA was achieved in a discharge prepared by rampdown of the Ohmic current. Modulation of microwave reflectometry signals at the fast wave frequency is being used to investigate fast wave propagation and damping. Additionally, rf pick-up probes on the internal boundary of the vessel provide a comparison with ray tracing codes, with dear evidence for a toroidally directed wave with antenna phasing set for current drive. There is some experimental evidence for fast wave absorption by energetic beam ions at high cyclotron harmonic resonances

  2. Fast wave current drive on DIII-D

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I.; Forest, C.B.; Ikezi, H.; Prater, R.; Baity, F.W.; Callis, R.W.; Cary, W.P.; Chiu, S.C.; Doyle, E.J.; Ferguson, S.W.; Hoffman, D.J.; Jaeger, E.F.; Kim, K.W.; Lee, J.H.; Lin-Liu, Y.R.; Murakami, M.; ONeill, R.C.; Porkolab, M.; Rhodes, T.L.; Swain, D.W.

    1996-01-01

    The physics of electron heating and current drive with the fast magnetosonic wave has been demonstrated on DIII-D, in reasonable agreement with theoretical modeling. A recently completed upgrade to the fast wave capability should allow full noninductive current drive in steady state advanced confinement discharges and provide some current density profile control for the Advanced Tokamak Program. DIII-D now has three four-strap fast wave antennas and three transmitters, each with nominally 2 MW of generator power. Extensive experiments have been conducted with the first system, at 60 MHz, while the two newer systems have come into operation within the past year. The newer systems are configured for 60 to 120 MHz. The measured FWCD efficiency is found to increase linearly with electron temperature as γ=0.4x10 18 T e0 (keV) [A/m 2 W], measured up to central electron temperature over 5 keV. A newly developed technique for determining the internal noninductive current density profile gives efficiencies in agreement with this scaling and profiles consistent with theoretical predictions. Full noninductive current drive at 170 kA was achieved in a discharge prepared by rampdown of the Ohmic current. Modulation of microwave reflectometry signals at the fast wave frequency is being used to investigate fast wave propagation and damping. Additionally, rf pick-up probes on the internal boundary of the vessel provide a comparison with ray tracing codes, with clear evidence for a toroidally directed wave with antenna phasing set for current drive. copyright 1996 American Institute of Physics

  3. Recent results from DIII-D and future plans

    International Nuclear Information System (INIS)

    Simonen, T.

    1992-01-01

    This paper summarizes recent DIII-D tokamak experimental results, describes new hardware being implemented to carry out the DIII-D 1990's tokamak research program, and discusses their implications for engineering designs for next generation tokamaks, such as ITER

  4. ITER vacuum vessel, in vessel components and plasma facing materials

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Enoeda, M.; Federici, G.

    2007-01-01

    Design of the NB ports including duct liners under heat loads of the neutral beams has been developed. Design of the in-wall shielding has been developed in more details considering the supporting structure and the assembly method. The ferromagnetic inserts have previously not been installed in the outboard midplane region due to irregularity caused by the tangential ports for NB injection. Due to this configuration, the maximum ripple is relatively large (∝1 %) in a limited region of the plasma and the toroidal field flux lines fluctuate ∝10 mm in the FW region. To avoid these problems, additional ferromagnetic inserts are to be installed in the equatorial port region. Detailed studies were carried out on the ITER vacuum vessel to define appropriate codes and standards in the context of the ITER licensing in France. A set of draft documents regarding the ITER vacuum vessel structural code were prepared including an RCC-MR Addendum for the ITER VV with justified exceptions or modifications. The main deviation from the base Code is the extensive use of UT in lieu of radiography for the volumetric examination of all one-side access welds of the outer shell and field joint. The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. Development of new beryllium materials is progressing in China and Russia. The ITER limiters will be installed in equatorial ports at two toroidal locations. The limiter plasma-facing surface protrudes ∝8 cm from the FW during the start-up and shutdown phase. In the new limiter concept, the limiters are retracted by ∝8 cm during the plasma flat top phase. This concept gives important advantages; (i) mitigation of the particle and heat loads due to disruptions, ELMs and blobs, (ii) improvement of the power coupling with the ICRH antenna

  5. Metastable beta limit in DIII-D

    International Nuclear Information System (INIS)

    La Haye, R.J.; Callen, J.D.; Gianakon, T.A.

    1997-06-01

    The long-pulse, slowly evolving single-null divertor (SND) discharges in DIII-D with H-mode, ELMs, and sawteeth are found to be limited significantly below (factor of 2) the predicted ideal limit β N = 4l i by the onset of tearing modes. The tearing modes are metastable in that they are explained by the neoclassical bootstrap current (high β θ ) destabilization of a seed island which occurs even if Δ' θ , there is a region of the modified Rutherford equation such that dw/dt > 0 for w larger than a threshold value; the plasma is metastable, awaiting the critical perturbation which is then amplified to the much larger saturated island. Experimental results from a large number of tokamaks indicate that the high beta operational envelope of the tokamak is well defined by ideal magnetohydrodynamic (MHD) theory. The highest beta values achieved have historically been obtained in fairly short pulse discharges, often <1-2 sawteeth periods and < 1-2 energy replacement times. The maximum operational beta in single-null divertor (SND), long-pulse discharges in DIII-D with a cross-sectional shape similar to the proposed ITER tokamak is found to be limited significantly below the threshold for ideal instabilities by the onset of resistive MHD instabilities

  6. Global Alfven Eigenmodes in DIII-D

    International Nuclear Information System (INIS)

    Turnbull, A.D.; Chu, M.S.; Strait, E.J.; Lao, L.L.; Greene, J.M.; Taylor, T.S.; Heidbrink, W.W.; Duong, H.; Chance, M.S.

    1992-06-01

    Global Alfven modes, such as the Toroidicity-Induced Alfven Eigenmode (TAE), pose a serious threat for strongly-heated tokamaks since they can result in saturation of the achievable beam β at moderate levels and they may also cause serious α-particle losses in future ignited devices. The DIII-D tokamak has a unique capability for study of the resonant excitation of these instabilities by energetic beam ions. TAE modes have now been observed in DIII-D over a wide range of operating conditions, including both circular cross-section and elongated (κ ∼ 1.8) discharges. Equilibrium reconstructions of several representative discharges, using all available external magnetic and internal profile data, have been done and analyzed in detail. The computed real mode frequencies of the TAE modes are in good agreement with the experimentally observed mode frequencies and differ significantly from the estimated kinetic ballooning mode frequencies. The TAE calculations include coupling to the Alfven and acoustic continuum branches of the MHD spectrum and generally indicate that the simplified circular cross-section, large aspect-ratio assumptions made in analytic calculations are poor approximations to the actual TAE mode structures. In particular, the global TAE modes are almost always coupled to one or more continuum branches by toroidicity, poloidal shaping, and finite β effects. Estimates of the various resonant excitation and damping mechanisms, including continuum damping, have been made and the total is found to be in reasonable agreement with the experimental threshold

  7. DIII-D research program progress

    Energy Technology Data Exchange (ETDEWEB)

    Stambaugh, R.D.

    1990-11-01

    A summary of highlights of the research on the DIII-D tokamak in the last two years is given. At low q, toroidal beta ({beta}{sub T}) has reached 11%. At high q, {epsilon}{beta}{sub p} has reached 1.8. DIII-D data extending from one regime to the other show the beta limit is at least {beta}{sub T}(%) {ge} 3.5 I/aB (MA, m, T). Prospects for using H-mode in future devices have been enhanced. The discovery of negative edge electric fields and associated turbulence suppression have become part of an emerging theory of H-mode. Long pulse (10 second) H-mode with impurity control has been demonstrated. Radial sweeping of the divertor strike points and gas puffing under the X-point have lowered peak divertor plate heat fluxes a factor of 3 and 2 respectively. T{sub i} = 17 keV has been reached in a hot ion H-mode. Electron cyclotron current drive (ECCD) has produced up to 70 kA of driven current. Program elements now beginning are fast wave current drive (FWCD) and an advanced divertor program (ADP). 38 refs., 10 figs.

  8. Upgraded divertor Thomson scattering system on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.; Taussig, D. A.; Boivin, R. L. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); McLean, A. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States)

    2016-11-15

    A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (T{sub e} in the range of 0.5 eV–2 keV, n{sub e} in the range of 5 × 10{sup 18}–1 × 10{sup 21} m{sup 3}) for both low T{sub e} in detachment and high T{sub e} measurement up beyond the separatrix.

  9. Disruption mitigation studies in DIII-D

    International Nuclear Information System (INIS)

    Taylor, P.L.; Kellman, A.G.; Evans, T.E.

    1999-01-01

    Data on the discharge behavior, thermal loads, halo currents, and runaway electrons have been obtained in disruptions on the DIII-D tokamak. These experiments have also evaluated techniques to mitigate the disruptions while minimizing runaway electron production. Experiments injecting cryogenic impurity killer pellets of neon and argon and massive amounts of helium gas have successfully reduced these disruption effects. The halo current generation, scaling, and mitigation are understood and are in good agreement with predictions of a semianalytic model. Results from killer pellet injection have been used to benchmark theoretical models of the pellet ablation and energy loss. Runaway electrons are often generated by the pellets and new runaway generation mechanisms, modifications of the standard Dreicer process, have been found to explain the runaways. Experiments with the massive helium gas puff have also effectively mitigated disruptions without the formation of runaway electrons that can occur with killer pellets

  10. Advanced tokamak research in DIII-D

    International Nuclear Information System (INIS)

    Greenfield, C M; Murakami, M; Ferron, J R

    2004-01-01

    Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with β N ≤ 3.5 and β T ≤ 3.6% sustained for up to 1 s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by

  11. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.; Buzhinskij, O.I.; Opimach, I.V.

    1998-08-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point (OSP) of two divertor plasma conditions: attached (T e > 40 eV) ELMing plasmas, and detached (T e 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y ≤ 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/exposure-year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates at the OSP of an attached plasma (∼ 10 microm/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  12. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.

    1998-05-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te 10 cm/year, even with incident heat flux 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D + ) ≤ 2.0 x 10 -3 ). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates (∼ 10 microm/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  13. Device for supporting the vacuum vessel of a thermonuclear device

    International Nuclear Information System (INIS)

    Sato, Hiroshi.

    1980-01-01

    Purpose: To hold a vacuum vessel securely at a predetermined position. Constitution: A vacuum vessel is supported on its one side to the standard mounting location of a support frame by way of a pin junction. The vacuum vessel is provided at its upper and lower positions with movable mounting portions, which are connected by way of connecting rods to fixed mounting locations on the upper and lower frames. The fixed mounting locations are disposed on a vertical plane including the axis of the torus center. This arrangement enables to hold even a large vacuum vessel at an exact predetermined position even under high temperature conditions without limiting the container's thermal expansion relative to the changes in temperature, thereby providing an extremely high rigidity against electromagnetic forces, earthquakes, etc. (Furukawa, Y.)

  14. Thermal-stress analysis and testing of DIII-D armor tiles

    International Nuclear Information System (INIS)

    Baxi, C.B.; Anderson, P.M.; Reis, E.E.; Smith, J.P.; Smith, P.D.; Croesmann, C.; Watkins, J.; Whitley, J.

    1987-10-01

    It is planned to install about 1500 new armor tiles in the DIII-D tokamak. The armor tiles currently installed in DIII-D are made by brazing Poco AXF-5Q graphite onto Inconel X-750 stock. A small percentage of these have failed by breakage of graphite. These failures were believed to be related to significant residual stress remaining in graphite after brazing. Hence, an effort was undertaken to improve the design with all-graphite tiles. Three criteria must be satisfied by the armor tiles and the hardware used to attach the tiles to the vessel walls: tiles should not structurally fail, peak tile temperature must be less than 2500 K, and peak vessel stresses must be below acceptable levels. A number of alternate design concepts were first analyzed with the two-dimensional finite element codes TOPAZ2D and NIKE2D. Promising designs were optimized for best parameters such as thicknesses, etc. The two best designs were further analyzed for thermal stresses with the three-dimensional codes P/THERMAL and P/STRESS. Prototype tiles of a number of materials were fabricated by GA and tested at the Plasma Materials Test Facility of the Sandia National Laboratory at Albuquerque. The tests simulated the heat flux and cooling conditions in DIII-D. This paper describes the 2-D and 3-D thermal stress analyses, the test results and logic which led to the selected design of the DIII-D armor tiles. 5 refs., 7 figs., 3 tabs

  15. DIII-D tokamak long range plan. Revision 3

    International Nuclear Information System (INIS)

    1992-08-01

    The DIII-D Tokamak Long Range Plan for controlled thermonuclear magnetic fusion research will be carried out with broad national and international participation. The plan covers: (1) operation of the DIII-D tokamak to conduct research experiments to address needs of the US Magnetic Fusion Program; (2) facility modifications to allow these new experiments to be conducted; and (3) collaborations with other laboratories to integrate DIII-D research into the national and international fusion programs. The period covered by this plan is 1 November 19983 through 31 October 1998

  16. Design and manufacturing of vacuum vessel of TPE-RX

    Energy Technology Data Exchange (ETDEWEB)

    Sago, H.; Kaguchi, H.; Orita, J.; Ishigami, Y. [Mitsubishi Heavy Industries Ltd., Kobe (Japan); Urata, K. [Mitsubishi Heavy Industries Ltd. (Japan). Nuclear Energy Systems Engineering Center; Hasegawa, M. [Mitsubishi Electric Co. (Japan). Nuclear Fusion Development; Yagi, Y.; Hirano, Y.; Shimada, T.; Sekine, S.; Sakakita, H. [Electrotechnical Lab. (Japan)

    1998-07-01

    Construction of a new, large reversed field pinch (RFP) machine called TPE-RX was complete at the end of 1997 as a successor of the previous TPE-1RM20 machine at the Electrotechnical Laboratory (ETL). RFP configuration has been successfully obtained in March 1998. This paper introduces structural design and manufacturing of the vacuum vessel of TPE-RX. The support positions were decided by structural analyses. The structural integrity of the vacuum vessel was evaluated by inelastic analyses. (author)

  17. Design and manufacturing of vacuum vessel of TPE-RX

    International Nuclear Information System (INIS)

    Sago, H.; Kaguchi, H.; Orita, J.; Ishigami, Y.; Urata, K.

    1998-01-01

    Construction of a new, large reversed field pinch (RFP) machine called TPE-RX was complete at the end of 1997 as a successor of the previous TPE-1RM20 machine at the Electrotechnical Laboratory (ETL). RFP configuration has been successfully obtained in March 1998. This paper introduces structural design and manufacturing of the vacuum vessel of TPE-RX. The support positions were decided by structural analyses. The structural integrity of the vacuum vessel was evaluated by inelastic analyses. (author)

  18. Depressurization as a means of leak checking large vacuum vessels

    International Nuclear Information System (INIS)

    Callis, R.W.; Langhorn, A.; Petersen, P.I.; Ward, C.; Wesley, J.

    1985-01-01

    A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, the vessel need only be depressurized 2 psi, thus allowing personnel inside to perform the sniffing operation. Second, the sniffing probe used (Leybold--Heraus ''Quick Test'') could detect a change in helium concentration as small as 100 ppb, which allows for faster scanning of the vessel inferior. Use of this technique to find an elusive 10 -3 Torrxl/s leak in the Doublet III tokamak vacuum vessel will be presented

  19. Metallurgical Bonding Development of V-4Cr-4Ti Alloy for the DIII-D Radiative Divertor Program

    International Nuclear Information System (INIS)

    Smith, J.P.; Johnson, W.R.; Trester, P.W.

    1998-01-01

    General Atomics (GA), in conjunction with the Department of Energy's (DOE) DIII-D Program, is carrying out a plan to utilize a vanadium alloy in the DIII-D tokamak as part of the DIII-D Radiative Divertor (RD) upgrade. The V-4Cr-4Ti alloy has been selected in the U.S. as the leading candidate vanadium alloy for fusion applications. This alloy will be used for the divertor fabrication. Manufacturing development with the V-4Cr-4Ti alloy is a focus of the DIII-D RD Program. The RD structure, part of which will be fabricated from V-4Cr-4Ti alloy, will require many product forms and types of metal/metal bonded joints. Metallurgical bonding methods development on this vanadium alloy is therefore a key area of study by GA. Several solid state (non-fusion weld) and fusion weld joining methods are being investigated. To date, GA has been successful in producing ductile, high strength, vacuum leak tight joints by all of the methods under investigation. The solid state joining was accomplished in air, i.e., without the need for a vacuum or inert gas environment to prevent interstitial impurity contamination of the V-4Cr-4Ti alloy

  20. Alternatives of ITER vacuum vessel support system

    International Nuclear Information System (INIS)

    Ohmori, Junji; Kitamura, Kazunori; Araki, Masanori; Ohno, Isamu; Shoji, Teruaki

    2002-07-01

    Optional designs of vacuum vessel (VV) support have been performed for the International Thermonuclear Experimental Reactor (ITER) to reduce stresses and buckling concern of the flexible plate structure in ITER-FDR. One of the optional designs is hanging type VV support concept that consists of top hanging supports at the top of VV and middle radial stoppers in the middle of outboard VV. The hanging supports are located at the toroidal field (TF) coil inboard top region (R∼5400 mm) using the narrow window space surrounded by a poloidal field coil (PF1) and TF coil. The radial stoppers are located inside TF coil cases in the TF coil outboard middle region (R∼9300 mm). The upper flange connection of the radial stoppers should slide in vertical direction to eliminate thermal stress produced by relative thermal displacement between VV wall and TF coil case. Both supports consist of flexible plates and are mounted on 18 locations in toroidal direction. The radial and toroidal reaction forces are shared with the hanging supports and the radial stoppers. However, the vertical force is sustained by only the hanging supports. The others are compressive type support concept that consists of nine VV supports located in alternate divertor port regions in toroidal direction. Two designs have been performed for the VV support concept. One is mounted on TF inter-coil structures (OIS) and the other is on cryostat ring. The compressive support on TF coil OIS is dependent on TF coil movement but that on cryostat is independent. In the optional designs, the bending stress due to the relative thermal displacement between TF coil and VV is classified to primary stress according to ASME Sec. III NF. The stress due to TF coil displacement is also considered as primary stress. The stress due to non-uniform temperature distribution of the flexible plate is classified to secondary stress. The preliminary structural assessments for the optional designs have been performed for all load

  1. ITER vacuum vessel design and electromagnetic analysis on in-vessel components

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

    1995-01-01

    Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

  2. DEMONSTRATION OF THE ITER IGNITION FIGURE OF MERIT AT q95>4 IN STATIONARY PLASMAS IN DIII-D

    International Nuclear Information System (INIS)

    WADE, M.R.; LUCE, T.C.; POLITZER, P.A.; FERRON, J.R.; HYATT, A.W.; SCOVILLE, J.T.; La HAYE, R.J.; KINSEY, J.E.; LASNIER, C.J.; MURAKAMI, M.; PETY, C.C.

    2002-01-01

    In order to maximize the probability of achieving ignition, the present International Thermonuclear Experimental Reactor (ITER) [1] design (as well as many of its predecessors) is based on operation at high plasma current. This constraint poses many significant engineering challenges, primarily related to the possibility of a sudden termination of the plasma current. Currents induced in the vessel and associated systems in such an event can lead to large forces, and runaway electrons may cause damage to the interior of the vacuum vessel. Present design methods (including those used for ITER) assume that the probability of experiencing such a major disruption increases with plasma current at fixed magnetic field and size. Because fusion performance is assumed to scale in a similar manner, reactor designs tend to seek a compromise between increased fusion performance and reduced susceptibility to disruptions, generally resulting in a design with q 95 ∼ 3.0. Discharges recently developed in the DIII-D tokamak offer a way to obtain equivalent fusion performance with more margin against disruption consequences, having obtained an ignition figure of merit comparable to the ITER baseline scenario with q 95 = 4.5. These discharges have been shown to be stationary on the thermal, resistive, and wall time scales and involve feedback control only of global quantities rather than profiles

  3. Plasma diagnostics for the DIII-D divertor upgrade (abstract)

    International Nuclear Information System (INIS)

    Hill, D.N.; Futch, A.; Buchenauer, D.; Doerner, R.; Lehmer, R.; Schmitz, L.; Klepper, C.C.; Menon, M.; Leikind, B.; Lippmann, S.; Mahdavi, M.A.; Schaffer, M.; Smith, J.; Salmonson, J.; Watkins, J.

    1990-01-01

    The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and H α measurements with TV cameras and fiber optics coupled to a high-resolution spectrometer

  4. System upgrades to the DIII-D facility

    International Nuclear Information System (INIS)

    Kellman, A.G.

    2007-01-01

    Major upgrades to the DIII-D facility have been performed that significantly enhance the capability of both the DIII-D device and the entire facility. The most significant of these include the rotation of a neutral beam line, installation of a new lower divertor, and a significant set of new and enhanced diagnostics. The upgrades and initial results are presented in this paper

  5. Automated Fault Detection for DIII-D Tokamak Experiments

    International Nuclear Information System (INIS)

    Walker, M.L.; Scoville, J.T.; Johnson, R.D.; Hyatt, A.W.; Lee, J.

    1999-01-01

    An automated fault detection software system has been developed and was used during 1999 DIII-D plasma operations. The Fault Identification and Communication System (FICS) executes automatically after every plasma discharge to check dozens of subsystems for proper operation and communicates the test results to the tokamak operator. This system is now used routinely during DIII-D operations and has led to an increase in tokamak productivity

  6. DIII-D Neutral Beam control system operator interface

    International Nuclear Information System (INIS)

    Harris, J.J.; Campbell, G.L.

    1993-10-01

    A centralized graphical user interface has been added to the DIII-D Neutral Beam (NB) control systems for status monitoring and remote control applications. This user interface provides for automatic data acquisition, alarm detection and supervisory control of the four NB programmable logic controllers (PLC) as well as the Mode Control PLC. These PLCs are used for interlocking, control and status of the NB vacuum pumping, gas delivery, and water cooling systems as well as beam mode status and control. The system allows for both a friendly user interface as well as a safe and convenient method of communicating with remote hardware that formerly required interns to access. In the future, to enable high level of control of PLC subsystems, complete procedures is written and executed at the touch of a screen control panel button. The system consists of an IBM compatible 486 computer running the FIX DMACS trademark for Windows trademark data acquisition and control interface software, a Texas Instruments/Siemens communication card and Phoenix Digital optical communications modules. Communication is achieved via the TIWAY (Texas Instruments protocol link utilizing both fiber optic communications and a copper local area network (LAN). Hardware and software capabilities will be reviewed. Data and alarm reporting, extended monitoring and control capabilities will also be discussed

  7. Operations Studies of the Gyrotrons on DIII-D

    Science.gov (United States)

    Storment, Stephen; Lohr, John; Cengher, Mirela; Gorelov, Yuri; Ponce, Dan; Torrezan, Antonio

    2017-10-01

    The gyrotrons are high power vacuum tubes used in fusion research to provide high power density heating and current drive in precisely localized areas of the plasma. Despite the increasing experience with both the manufacture and operation of these devices, individual gyrotrons with similar design and manufacturing processes can exhibit important operational differences in terms of generated rf power, efficiency and lifetime. This report discusses differences in the performance of several gyrotrons in operation at DIII-D and presents the results of a series of measurements that could lead to improved the performance of single units based on a better understanding of the causes of these differences. The rf power generation efficiency can be different from gyrotron to gyrotron. In addition, the power loading of the collector can feature localized hot spots, where the collector can locally be close to the power deposition limits. Measurements of collector power loading provide maps of the power deposition and can provide understanding of the effect of modulation of the output rf beam on the total loading, leading to improved operational rules increasing the safety margins for the gyrotrons under different operational scenarios. Work supported by US DOE under DE-FC02-04ER54698.

  8. Baking of SST-1 vacuum vessel modules and sectors

    International Nuclear Information System (INIS)

    Pathan, Firozkhan S; Khan, Ziauddin; Yuvakiran, Paravastu; George, Siju; Ramesh, Gattu; Manthena, Himabindu; Shah, Virendrakumar; Raval, Dilip C; Thankey, Prashant L; Dhanani, Kalpesh R; Pradhan, Subrata

    2012-01-01

    SST-1 Tokamak is a steady state super-conducting tokamak for plasma discharge of 1000 sec duration. The plasma discharge of such long time duration can be obtained by reducing the impurities level, which will be possible only when SST-1 vacuum chamber is pumped to ultra high vacuum. In order to achieve UHV inside the chamber, the baking of complete vacuum chamber has to be carried out during pumping. For this purpose the C-channels are welded inside the vacuum vessel. During baking of vacuum vessel, these welded channels should be helium leak tight. Further, these U-channels will be in accessible under operational condition of SST-1. So, it will not possible to repair if any leak is developed during experiment. To avoid such circumstances, a dedicated high vacuum chamber is used for baking of the individual vacuum modules and sectors before assembly so that any fault during welding of the channels will be obtained and repaired. This paper represents the baking of vacuum vessel modules and sectors and their temperature distribution along the entire surface before assembly.

  9. Disruption Physics and Mitigation on DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Humphreys, D.A.; Kellman, A.G.

    2005-01-01

    The contributions of the DIII-D tokamak toward the understanding and control of disruptions are reviewed. Disruptions are found to be deterministic, and the underlying causes of disruption can therefore be predicted and avoided. With sufficiently rapid detection, possible damage from disruptions can be mitigated using an understanding of disruption phenomenology and plasma physics. Regimes of high β are readily available in DIII-D and provide access to relatively high energy density disruptions, despite DIII-D's moderate magnetic field and size. DIII-D, with all-graphite wall armor and wall conditioning between discharges, has proven highly resilient to the deleterious effects that disruptions can have on plasma operations. Simultaneously, exploitation and adaptation of DIII-D's extensive core and edge plasma diagnostic set have allowed for unique plasma measurements during disruptions. These measurements have tied into the development of several physical models used to understand aspects of disruptions, such as magnetohydrodynamic growth at the disruption onset, radiation energy balance through the thermal quench, and halo currents during the current quench. Based on this fundamental understanding, DIII-D has developed techniques to mitigate the harmful effects of disruptions by radiative dissipation of the plasma energy and extrapolated these techniques for possible use on larger devices like ITER

  10. Control of the resistive wall mode with internal coils in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Okabayashi, M.; Bialek, J.; Bondeson, A.; Chance, M.S.; Chu, M.S.; Garofalo, A.M.; Hatcher, R.; In, Y.; Jackson, G.L.; Jayakumar, R.J.; Jensen, T.H.; Katsuro-Hopkins, O.; Haye, R.J. La; Liu, Y.Q.; Navratil, G.A.; Reimerdes, H.; Scoville, J.T.; Strait, E.J.; Takechi, M.; Turnbull, A.D.; Gohil, P.; Kim, J.S.; Makowski, M.A.; Manickam, J.; Menard, J.

    2005-01-01

    Internal coils, 'I-Coils', were installed inside the vacuum vessel of the DIII-D device to generate non-axisymmetric magnetic fields to act directly on the plasma. These fields are predicted to stabilize the resistive wall mode (RWM) branch of the long-wavelength external kink mode with plasma beta close to the ideal wall limit. Feedback using these I-Coils was found to be more effective as compared to using external coils located outside the vacuum vessel. Locating the coils inside the vessel allows for a faster response and the coil geometry also allows for better coupling to the helical mode structure. Initial results were reported previously (Strait E.J. et al 2004 Phys. Plasmas 11 2505). This paper reports on results from extended feedback stabilization operations, achieving plasma parameters up to the regime of C β ∼ 1.0 and open loop growth rates of γ open τ w ∼ 25 where the RWM was predicted to be unstable with only the 'rotational viscous stabilization mechanism'. Here C β ∼ (β - β no-wall.limit )/(β ideal.wall.limit - β no-wall.limit ) is a measure of the beta relative to the stability limits without a wall and with a perfectly conducting wall, and τ w is the resistive flux penetration time of the wall. These feedback experimental results clarified the processes of dynamic error field correction and direct RWM stabilization, both of which took place simultaneously during RWM feedback stabilization operation. MARS-F modelling provides a critical rotation velocity in reasonable agreement with the experiment and predicts that the growth rate increases rapidly as rotation decreases below the critical. The MARS-F code also predicted that for successful RWM magnetic feedback, the characteristic time of the power supply should be limited to a fraction of the growth time of the targeted RWM. The possibility of further improvements in the presently achievable range of operation of feedback gain values is also discussed

  11. Density limit studies on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge National Lab., TN (United States); Mahdavi, M.A.; Petrie, T.W. [General Atomics, San Diego, CA (United States)] [and others

    1998-08-01

    The authors have studied the processes limiting plasma density and successfully achieved discharges with density {approximately}50% above the empirical Greenwald density limit with H-mode confinement. This was accomplished by density profile control, enabled through pellet injection and divertor pumping. By examining carefully the criterion for MARFE formation, the authors have derived an edge density limit with scaling very similar to Greenwald scaling. Finally, they have looked in detail at the first and most common density limit process in DIII-D, total divertor detachment, and found that the local upstream separatrix density (n{sub e}{sup sep,det}) at detachment onset (partial detachment) increases with the scrape-off layer heating power, P{sub heat}, i.e., n{sub e}{sup sep,det} {approximately} P{sub heat}{sup 0.76}. This is in marked contrast to the line-average density at detachment which is insensitive to the heating power. The data are in reasonable agreement with the Borass model, which predicted that the upstream density at detachment would increase as P{sub heat}{sup 0.7}.

  12. Density limit studies on DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Mahdavi, M.A.; Petrie, T.W.

    1998-08-01

    The authors have studied the processes limiting plasma density and successfully achieved discharges with density ∼50% above the empirical Greenwald density limit with H-mode confinement. This was accomplished by density profile control, enabled through pellet injection and divertor pumping. By examining carefully the criterion for MARFE formation, the authors have derived an edge density limit with scaling very similar to Greenwald scaling. Finally, they have looked in detail at the first and most common density limit process in DIII-D, total divertor detachment, and found that the local upstream separatrix density (n e sep,det ) at detachment onset (partial detachment) increases with the scrape-off layer heating power, P heat , i.e., n e sep,det ∼ P heat 0.76 . This is in marked contrast to the line-average density at detachment which is insensitive to the heating power. The data are in reasonable agreement with the Borass model, which predicted that the upstream density at detachment would increase as P heat 0.7

  13. Alfven Eigenmode Control in DIII-D

    Science.gov (United States)

    Hu, W.; Olofsson, E.; Welander, A.; van Zeeland, M.; Collins, C.; Heidbrink, W.

    2017-10-01

    Alfven eigenmodes (AE) driven by fast ions from neutral beam and ion cyclotron heating are common in present day tokamak plasmas and are expected to be destabilized by alpha particles in future burning plasma experiments. Because these waves have been shown to cause loss and redistribution of fast ions which can impact plasma performance and potentially device integrity, developing control techniques for AEs is of paramount importance. In the DIII-D plasma control system, spectral analysis of real-time ECE data is used as a monitor of AE amplitude, frequency, and location. These values are then used for feedback control of the neutral beam power to control Alfven waves and reduce fast ion loss. This work describes tests of AE control experiments in the current ramp up phase, during which multiple Alfven eigenmodes are typically unstable and fast ion confinement is degraded significantly. Comparisons of neutron emission and confined fast ion profiles with and without active AE control will be made. Work supported by the U.S. Dept. of Energy under Award Number DE-FC02-04ER54698.

  14. The long range DIII-D plan

    International Nuclear Information System (INIS)

    Simonen, T.C.

    1994-01-01

    The mission of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. The National Energy Strategy calls for the development of magnetic fusion as an energy options with operation of a DEMO by 2025. The DEMO will be based on nuclear technology demonstrated in ITER and the physics and engineering database established in magnetic fusion facilities during the next two decades. On the present path, based on extrapolation of current conventional operating modes, ITER is twice as large as Joint European Tokamak (JET), and DEMO, using the same logic, will be even larger. However, successful development of advanced tokamak (AT) operating modes could open the way for significantly improved confinement and stability, leading to a smaller, more commercially attractive DEMO, provided new divertor concepts are developed to handle the accompanying high divertor power density. A smaller and lower cost DEMO opens up the possibility that multiple nations, utilities, and industries could build DEMOs simultaneously and, therefore, more rapidly optimize the tokamak for commercialization

  15. Vacuum vessel for the tandem Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Gerich, J.W.

    1986-01-01

    In 1980, the US Department of Energy gave the Lawrence Livermore National Laboratory approval to design and build a tandem Mirror Fusion Test Facility (MFTF-B) to support the goals of the National Mirror Program. We designed the MFTF-B vacuum vessel both to maintain the required ultrahigh vacuum environment and to structurally support the 42 superconducting magnets plus auxiliary internal and external equipment. During our design work, we made extensive use of both simple and complex computer models to arrive at a cost-effective final configuration. As part of this work, we conducted a unique dynamic analysis to study the interaction of the 32,000-tonne concrete-shielding vault with the 2850-tonne vacuum vessel system. To maintain a vacuum of 2 x 10 -8 torr during the physics experiments inside the vessel, we designed a vacuum pumping system of enormous capacity. The vacuum vessel (4200-m 3 internal volume) has been fabricated and erected, and acceptance tests have been completed at the Livermore site. The rest of the machine has been assembled, and individual systems have been successfully checked. On October 1, 1985, we began a series of integrated engineering tests to verify the operation of all components as a complete system

  16. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    Mardani, M.; Amrollahi, R.; Koohestani, S.

    2012-01-01

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  17. Integration of ITER in-vessel diagnostic components in the vacuum vessel

    International Nuclear Information System (INIS)

    Encheva, A.; Bertalot, L.; Macklin, B.; Vayakis, G.; Walker, C.

    2009-01-01

    The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement's deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence.

  18. Testing program for burning plasma experiment vacuum vessel bolted joint

    International Nuclear Information System (INIS)

    Hsueh, P.K.; Khan, M.Z.; Swanson, J.; Feng, T.; Dinkevich, S.; Warren, J.

    1992-01-01

    As presently designed, the Burning Plasma Experiment vacuum vessel will be segmentally fabricated and assembled by bolted joints in the field. Due to geometry constraints, most of the bolted joints have significant eccentricity which causes the joint behavior to be sensitive to joint clamping forces. Experience indicates that as a result of this eccentricity, the joint will tend to open at the side closest to the applied load with the extent of the opening being dependent on the initial preload. In this paper analytical models coupled with a confirmatory testing program are developed to investigate and predict the non-linear behavior of the vacuum vessel bolted joint

  19. Vacuum vessel of thermonuclear device and manufacturing method thereof

    International Nuclear Information System (INIS)

    Kurita, Genichi; Nagashima, Keisuke; Uchida, Takaho; Shibui, Masanao; Ebisawa, Katsuyuki; Nakagawa, Satoshi.

    1997-01-01

    The present invention provides a vacuum vessel of a thermonuclear device using, as a material of a plasma vacuum vessel, a material to be less activated and having excellent strength as well as a manufacturing method thereof. Namely, the vacuum vessel is made of titanium or a titanium alloy. In addition, a liner layer comprising a manganese alloy, nickel alloy, nickel-chromium alloy or aluminum or aluminum alloy is formed. With such a constitution, the wall substrate made of titanium or a titanium alloy can be isolated by the liner from hydrogen or plasmas. As a result, occlusion of hydrogen to titanium or the titanium alloy can be prevented thereby enabling to prevent degradation of the material of the wall substrate of the vacuum vessel. In addition, since the liner layer has relatively high electric resistance, a torus circumferential resistance value required for plasma ignition can be ensured by using it together with the vessel wall made of titanium alloy. (I.S.)

  20. A remote control room at DIII-D

    International Nuclear Information System (INIS)

    Abla, G.; Schissel, D.P.; Penaflor, B.G.; Wallace, G.

    2008-01-01

    This paper describes a remote control room built at DIII-D to support remote participation activities of DIII-D research staff. In order to create a persistent, efficient, and reliable remote participation environment for DIII-D scientists, a remote control room has been built in a 640-ft 2 dedicated area. The purpose of this room is to experiment and define a remote control room framework that can facilitate the remote participation needs of current and future fusion experiments such as ITER. A variety of hardware equipment has been installed and several remote participation and collaboration technologies have been deployed. Objectivity and practical consideration has been the key while designing the room and deploying the technologies. Although, the DIII-D remote control room is still a work in progress and new software tools are being implemented, it has been already useful for a number of international remote participation activities. For example, it has been used for remote support of the EAST Tokamak in China during the start up operation and proven effective for other collaborative experiment activities. The description of the remote control room design is given along with technologies deployed for remote collaboration needs. We will also discuss our recent experiences involving the DIII-D remote control room as well as future plans for improvements

  1. Vacuum vessels for the LHC magnets arrive at CERN

    CERN Multimedia

    2001-01-01

    The first batch of pre-series vacuum vessels for the LHC dipole magnets has just been delivered to CERN. The vessels are components of the cryostats and will provide the thermal insulation for the superconducting magnets. The first batch of vacuum vessels for the LHC dipole magnets with the team taking part at CERN in ordering and installing them. Left to right : Claude Hauviller, Monique Dupont, Lloyd Williams, Franck Gavin, Alain Jacob, Christophe Vuitton, Davide Bozzini, Laure Sandri, Mikael Sjoholm and André de Saever. In 2006 all that will be seen of the LHC superconducting dipoles in the LHC tunnel will be a line of over 1230 blue cylindrical vacuum vessels. Ten vessels, each weighing 4 tonnes, are already at CERN. On 6 July the first batch of pre-series vessels reached the Lab-oratory from the firm SIMIC Spa whose works are near Savona in north-western Italy. Despite appearances, these 15-metre long, 1-metre diameter blue tubes are much more sophisticated than sections of a run-of-the-mill...

  2. Effect of variation in equilibrium shape on ELMing H-mode performance in DIII-D diverted plasmas

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Osborne, T.H.; Petrie, T.W.

    2001-01-01

    The changes in the performance of the core, pedestal, scrape-off-layer (SOL), and divertor plasmas as a result of changes in triangularity, δ, up/down magnetic balance, and secondary divertor volume were examined in shape variation experiments using ELMing H mode plasmas on DIII-D. In moderate density, unpumped plasmas, high δ∼0.7 increased the energy in the H mode pedestal and the global energy confinement of the core, primarily due to an increase in the margin by which the edge pressure gradient exceeded the value which would have been expected had it been limited by infinite-n ideal ballooning modes. In addition, a nearly balanced double-null (DN) shape was effective for sharing the peak heat flux in the divertor in these attached plasmas. For detached plasmas good heat flux sharing was obtained for a substantial range of unbalanced DN shapes. Finally, the presence of a second X-point in unbalanced DN shapes did not degrade the plasma performance if it was sufficiently far inside the vacuum vessel. These results indicate that a high δ unbalanced DN shape has some advantages over a single null shape for future high power tokamak operation. (author)

  3. The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL

    2017-08-01

    Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

  4. TSC plasma halo simulation of a DIII-D vertical displacement episode

    International Nuclear Information System (INIS)

    Sayer, R.O.; Peng, Y.K.M.; Jardin, S.C.

    1993-01-01

    A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench and current quench. With a suitable halo surrounding the main plasma, the TSC predictions are in good agreement with experimental results for the plasma current decay, plasma trajectory, toroidal and poloidal vessel currents, and for the magnetic probe and flux loop values for the entire VDE. Simulations with no plasma halo yield much faster vertical motion and significantly worse agreement with the magnetics and flux loop data than do halo simulations. (author). 12 refs, 13 figs

  5. Welding distortion control in double walled KSTAR vacuum vessel fabrication

    International Nuclear Information System (INIS)

    Oh, D. W.; Lee, G. T.; Kim, H. K.; Yang, H. L.; Bak, J. S.

    2004-01-01

    The KSTAR(Korea Superconducting Tokamak Advanced Research) vacuum vessel is designed to be a double walled structure made of 12mm thick 316LN stainless steel with a D shaped cross-section about 4 m height. Vacuum vessel was pre-fabricated in two parts, 180 degree and 157.5 degree sectors in toroidal direction to meet the transportation purpose. These two parts have to be welded on site with ±2mm allowable fabrication tolerances. 1/3 scaled mock-up model was used to estimate the welding distortion and to ensure the weld quality of vacuum vessel. Gas Tungsten Arc Welding(GTAW), which has been approved by procedure qualification test, was used during mock-up test and vacuum vessel site fabrication. Welding distortion could be managed by allowing for distortion in opposite direction, by applying high restraint using lots of strong backs, by controlling the welding heat input with symmetrical welding sequence. The integrity of the site welding joint was assured by radiographic test, ultrasonic test and leak test with helium detecting method

  6. Analysis of effective electrical parameters for CFETR vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xufeng; Xu, Weiwei, E-mail: wwxu@ipp.ac.cn; Du, Shuangsong; Zheng, Jinxing

    2016-11-15

    Highlights: • The eddy current distribution and variation of CFETR vacuum vessel during plasma disruption have been calculated. • Effective electrical parameters can be derived from the eddy current characters. • The method for eddy current and effective electrical parameters is suit for the complex shell with arbitrary shape. - Abstract: The electrical parameters of CFETR (China Fusion Engineering Test Reactor) vacuum vessel are very important to the design of control system and power supply system. Effective electrical parameters are relevant to the dynamic of eddy current. For complex structure, the distribution of eddy current can’t be obtained by analytical form. A method is presented to solve the eddy current of the vacuum vessel in this paper. The effective electrical parameters can be got from the eddy current distribution and variation. The time constant of the CFETR vacuum vessel is derived from the decay characteristics of the eddy current. And the effective resistance and inductance can be derived from the viewpoint of energy for a certain distribution of eddy current.

  7. Advances in integrated plasma control on DIII-D

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Hahn, S.H.; Humphreys, D.A.; In, Y.; Johnson, R.D.; Kim, J.S.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Welander, A.S.; Xiao, B.

    2007-01-01

    The DIII-D advanced tokamak physics program requires extremely high performance from the DIII-D plasma control system, including simultaneous accurate regulation of plasma shape, stored energy, density and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another, simulation of control action against off-line and actual machine control platforms, and optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of new model-based plasma controllers on DIII-D. We also describe the development of the control software, hardware, and model-based control algorithms for the superconducting EAST and KSTAR tokamaks

  8. Enhanced DIII-D Data Management Through a Relational Database

    Science.gov (United States)

    Burruss, J. R.; Peng, Q.; Schachter, J.; Schissel, D. P.; Terpstra, T. B.

    2000-10-01

    A relational database is being used to serve data about DIII-D experiments. The database is optimized for queries across multiple shots, allowing for rapid data mining by SQL-literate researchers. The relational database relates different experiments and datasets, thus providing a big picture of DIII-D operations. Users are encouraged to add their own tables to the database. Summary physics quantities about DIII-D discharges are collected and stored in the database automatically. Meta-data about code runs, MDSplus usage, and visualization tool usage are collected, stored in the database, and later analyzed to improve computing. Documentation on the database may be accessed through programming languages such as C, Java, and IDL, or through ODBC compliant applications such as Excel and Access. A database-driven web page also provides a convenient means for viewing database quantities through the World Wide Web. Demonstrations will be given at the poster.

  9. The design and fabrication of a toroidally continuous cryocondensation pump for the DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.J.; Schaubel, K.M.; Menon, M.M.

    1991-11-01

    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 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. Key design considerations were: how to accommodate the temperature differences between the various components, developing low heat leak paths for the various supports, and maintaining electrical insulation in a low pressure environment in the presence of induced voltage spikes. A single point ground for the system was used to limit disruption induced currents and the resulting electro-mechanical forces on the pump. 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

  10. New DIII-D tokamak plasma control system

    International Nuclear Information System (INIS)

    Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.

    1992-09-01

    A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter

  11. Fabrication and installation of the DIII-D radiative divertor structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.

    1997-11-01

    Phase 1A of the Radiative Divertor Program (RDP) is now installed in the DIII-D tokamak located at General Atomics. This hardware was added to enhance both the Divertor and Advanced Tokamak research elements of the DIII-D program. This installation consists of a divertor baffle enveloping a cryocondensation pump at the upper outer divertor target of DIII-D. The divertor baffle consists of two toroidally continuous Inconel 625 water-cooled rings and a toroidal array of discontinuous radiatively-cooled plates. The water-cooled rings are each comprised of four quadrants, mechanically formed, chem.-milled, and resistance and TIG welded Inconel 625 panels. The supports attaching the panels to the vessel wall are designed to accommodate the differential thermal expansion between the rings and vessel during bake and to react the electromagnetic loads induced during disruptions. They are made from either Inconel 625 or Inconel 718 depending on the stress levels predicted in Finite Element Analysis. Gas seals are designed to limit the leakage from the baffle chamber back to the core plasma to 2,500 ell/s and incorporate plasma sprayed alumina to minimize currents flowing through them. The bulk of the water-cooled ring fabrication was performed by a vendor, however, the final machining of penetrations in the conical ring for diagnostic access was performed in-house using a unique machining configuration. This configuration, and the machining of the diagnostic cutouts is described. Graphite tiles were machined from ATJ graphite to form a smooth plasma-facing surface. The installation of all divertor components required only four weeks

  12. Structural design and manufacturing of TPE-RX vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y. [Mitsubishi Heavy Ind. Ltd., Kobe (Japan); Urata, K.; Kudough, F. [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan); Hasegawa, M.; Oyabu, I. [Mitsubishi Electric Co., Tokyo (Japan); Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H. [Electrotechnical Laboratory, Tsukuba (Japan)

    1999-10-01

    TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

  13. Structural design and manufacturing of TPE-RX vacuum vessel

    International Nuclear Information System (INIS)

    Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y.; Urata, K.; Kudough, F.; Hasegawa, M.; Oyabu, I.; Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H.

    1999-01-01

    TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

  14. New baking system for the RFX vacuum vessel

    International Nuclear Information System (INIS)

    Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G.

    1996-01-01

    A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320 degree C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs

  15. New baking system for the RFX vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G. [Universita di Padova (Italy)

    1996-12-31

    A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320{degree}C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs.

  16. FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

    International Nuclear Information System (INIS)

    O'NEIL, RC; STAMBAUGH, RD

    2002-01-01

    OAK A271 FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT. The Radiative Divertor Project originated in 1993 when the DIII-D Five Year Plan for the period 1994--1998 was prepared. The Project Information Sheet described the objective of the project as ''to demonstrate dispersal of divertor power by a factor of then with sufficient diagnostics and modeling to extend the results to ITER and TPX''. Key divertor components identified were: (1) Carbon-carbon and graphite armor tiles; (2) The divertor structure providing a gas baffle and cooling; and (3) The divertor cryopumps to pump fuel and impurities

  17. Cooperative program on DIII-D (FY93)

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1994-01-01

    This is a proposal to continue support of the authors cooperative research program on DIII-D, under Department of Energy contract DE-FG03-89ER51116. The proposal describes work carried out recently in support of DIII-D data analysis and modeling, with a focus on divertors, edge physics and transport phenomena linking edge and core physics. Proposed work will continue to focus on edge physics, instabilities, the further development of codes to model the plasma, and data analysis in support of related experimental work

  18. Design of the divertor Thomson scattering system on DIII-D

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Foote, J.H.; Nilson, D.G.; Rice, B.W.

    1994-05-01

    Local measurements of n e and T e in the divertor region are necessary for a more complete understanding of divertor physics. We have designed an extension to the existing multipulse Thomson scattering system to measure n e in the range 5 x 10 18 to 5 x 10 20 m -3 and T e 5--500 eV, with 1 cm resolution from 1--21 cm above the floor of the DIII-D vessel, in the region of the X-point for lower single-null diverted plasmas. One of the existing 8, 20 Hz, ND:YAG lasers will be redirected to a separate vertical port, and viewed radially with a specially designed, f/6.8 lens. Fiber optics carry the light to additional polychromators whose interference filters have been optimized for low T e measurements. Other aspect of the system, including the beam path to the vessel, polychromator design, real time data acquisition, laser control, calibration facility, and DIII-D timing and data acquisition interface will be shared with the existing multipulse Thomson system. An in-situ laser alignment monitor will provide alignment information for each laser pulse

  19. Wall conditioning and plasma surface interactions in DIII-D

    International Nuclear Information System (INIS)

    Jackson, G.L.; Petersen, P.I.; Schaffer, M.S.; Taylor, P.L.; Taylor, T.S.; Doyle, B.L.; Walsh, D.S.; Hill, D.N.; Hsu, W.L.; Winter, J.

    1990-09-01

    Wall conditioning is used in DIII-D for both reduction of impurity influxes and particle control. The methods used include: baking, pulsed discharge cleaning, hydrogen glow cleaning, helium and neon glow conditioning, and carbonization. Helium glow wall conditioning applied before every tokamak discharge has been effective in impurity removal and particle control and has significantly expanded the parameter space in which DIII-D operates to include limiter and ohmic H-mode discharges and higher β T at low q. The highest values of divertor plasma current (3.0 MA) and stored energy (3.6 MJ) and peaked density profiles in H-mode discharges have been observed after carbonization. Divertor physics studies in DIII-D include sweeping the X-point to reduce peak heat loads, measurement of particle and heat fluxes in the divertor region, and erosion studies. The DIII-D Advanced Divertor has been installed and bias and baffle experiments will begin in the fall of 1991. 15 refs., 4 figs

  20. The DIII-D Computing Environment: Characteristics and Recent Changes

    International Nuclear Information System (INIS)

    McHarg, B.B. Jr.

    1999-01-01

    The DIII-D tokamak national fusion research facility along with its predecessor Doublet III has been operating for over 21 years. The DIII-D computing environment consists of real-time systems controlling the tokamak, heating systems, and diagnostics, and systems acquiring experimental data from instrumentation; major data analysis server nodes performing short term and long term data access and data analysis; and systems providing mechanisms for remote collaboration and the dissemination of information over the world wide web. Computer systems for the facility have undergone incredible changes over the course of time as the computer industry has changed dramatically. Yet there are certain valuable characteristics of the DIII-D computing environment that have been developed over time and have been maintained to this day. Some of these characteristics include: continuous computer infrastructure improvements, distributed data and data access, computing platform integration, and remote collaborations. These characteristics are being carried forward as well as new characteristics resulting from recent changes which have included: a dedicated storage system and a hierarchical storage management system for raw shot data, various further infrastructure improvements including deployment of Fast Ethernet, the introduction of MDSplus, LSF and common IDL based tools, and improvements to remote collaboration capabilities. This paper will describe this computing environment, important characteristics that over the years have contributed to the success of DIII-D computing systems, and recent changes to computer systems

  1. Simple catalytic cell for restoring He leak detector sensitivity on vacuum systems with high D2 backgrounds

    International Nuclear Information System (INIS)

    Busath, J.; Chiu, H.K.

    1998-12-01

    The DIII-D National Fusion Facility at General Atomics focuses on plasma physics and fusion energy science. The DIII-D tokamak is a 35 m 3 toroidal vacuum vessel with over 200 ports for diagnostic instrumentation, cryogenics, microwave heating, and four large neutral beam injectors. Maintaining vacuum in the 10 -8 Torr range is crucial for producing high performance plasma discharges. He leak checking the DIII-D tokamak and the neutral beamlines has historically been difficult. D 2 is used as the fuel gas in most plasma discharges and neutral beams. After plasma operations, D 2 out-gassing from the torus walls and internal beamline components can exceed 10 -4 std cc/s. The mass of the D 2 molecule (4.028 u) is indistinguishable from that of the He atom (4.003 u) to a standard mass spectrometer leak detector. High levels of D 2 reduce leak detector sensitivity and effectively mask the He trace gas signal rendering normal leak checking techniques ineffective. A simple apparatus was developed at GA to address these problems. It consists of a palladium based catalyst cell and associated valves and piping placed in series with the leak detector. This reduces the D 2 throughput by a factor greater than 10,000, restoring leak detector sensitivity. This paper will briefly discuss the development of the cell, the physical processes involved, the tests performed to quantify and optimize the processes, and the operational results at DIII-D

  2. Manufacture, testing and assembly preparation of the JET vacuum vessel

    International Nuclear Information System (INIS)

    Arbez, J.; Baeumel, S.; Dean, J.R.; Duesling, G.; Froger, C.; Hemmerich, J.L.; Walravens, M.; Walter, K.; Winkel, T.

    1983-01-01

    To reach the target pressure of 10 -9 mbar, JET's double-walled Inconel vacuum vessel is being manufactured and assembled in clean conditions and with meticulous leak detection. Each octant (1/8 of the torus) is baked in an oven to 520 0 C and leak tested at 350 0 C to reveal leaks as small as 10 -9 mbar l/s, which are repaired. In service the vessel will be baked periodically to 500 0 C by CO 2 passing between its walls. The single-walled ports will be electrically heated. (author)

  3. Fabrication progress of the ITER vacuum vessel sector in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.C., E-mail: bckim@nfri.re.kr [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Y.J.; Hong, K.H.; Sa, J.W.; Kim, H.S.; Park, C.K.; Ahn, H.J.; Bak, J.S.; Jung, K.J. [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Park, K.H.; Roh, B.R.; Kim, T.S.; Lee, J.S.; Jung, Y.H.; Sung, H.J.; Choi, S.Y.; Kim, H.G.; Kwon, I.K.; Kwon, T.H. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of)

    2013-10-15

    Highlights: ► Fabrication of ITER vacuum vessel sector full scale mock-up to develop fabrication procedures. ► The welding and nondestructive examination techniques conform to RCC-MR. ► The preparation of real manufacturing of ITER vacuum vessel sector. -- Abstract: As a participant of ITER project, ITER Korea has to supply two ITER vacuum vessel sectors (Sector no. 6, no. 1) of total nine ITER VV sectors. After the procurement arrangement with ITER Organization, ITER Korea made the contract with Hyundai Heavy Industries (HHI) for fabrication of two sectors. Then the start of the manufacturing design was initiated from January 2010. HHI made three real scale R and D mock-ups to verify the critical fabrication feasibility issues on electron beam welding, 3D forming, welding distortion and achievable tolerances. The documentation according to IO and the French nuclear safety regulation requirement, the qualification of welding and nondestructive examination procedures conform to RCC-MR 2007 were proceed in parallel. The mass production of raw material was done after receiving ANB (agreed notified body) verification of product/parts and shop qualification. The manufacturing drawing, manufacturing and inspection plan of VV sector with supporting fabrication procedures are also verified by ANB, accordingly the first cutting and forming of plates for VV sector fabrication started from February 2012. This paper reports the latest fabrication progress of ITER vacuum vessel Sector no. 6 that will be assembled as the first sector in the ITER pit. The overall fabrication route, R and D mock-up fabrication results with forming and welding distortion analysis, qualification status of welding and nondestructive examination (NDE) are also presented.

  4. ITER vacuum vessel structural analysis completion during manufacturing phase

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F. [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Caixas, J.; Fernandez, E.; Zarzalejos, J.M. [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Kim, H.-S.; Kim, Y.G. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Privalova, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); and others

    2016-11-01

    Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

  5. ATLAS Supplier Award for the ECT Vacuum Vessels

    CERN Multimedia

    Jenni, P

    On 12 February the Netherlands firm Schelde Exotech was awarded the ATLAS Supplier Award for the construction of the two vacuum vessels for the ATLAS End- Cap Toroid (ECT) magnets. ATLAS Supplier Award ceremonies have now become something of a tradition. For the third consecutive year, ATLAS has given best supplier awards for the most exceptional contributors to the construction of the detector. The Netherlands firm Schelde Exotech has just received the award for the construction of the two vacuum vessels for the ECTs. With a diameter of 11 metres and a volume of 550 cubic metres, the ECT vacuum vessels are obviously impressive in scale. They consist of large aluminium plates and a stainless steel central bore tube. In order to obtain the required undulations, the firm had to develop a special assembly and welding technique. Despite the chambers' imposing size, a very high degree of precision has been achieved in their geometry. Moreover, the chambers, which were delivered in July 2002 to CERN, were built i...

  6. ITER vacuum vessel structural analysis completion during manufacturing phase

    International Nuclear Information System (INIS)

    Martinez, J.-M.; Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F.; Caixas, J.; Fernandez, E.; Zarzalejos, J.M.; Kim, H.-S.; Kim, Y.G.; Privalova, E.

    2016-01-01

    Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

  7. ITER vacuum vessel dynamic stress analysis of a disruption

    International Nuclear Information System (INIS)

    Riemer, B.W.; Conner, D.L.; Strickler, D.J.; Williamson, D.E.

    1994-01-01

    Dynamic stress analysis of the International Thermonuclear Experimental Reactor vacuum vessel loaded by disruption forces was performed. The deformation and stress results showed strong inertial effects when compared to static analyses. Maximum stress predicted dynamically was 300 MPa, but stress shown by static analysis from loads at the same point in time reached only 80 MPa. The analysis also provided a reaction load history in the vessel's supports which is essential in evaluating support design. The disruption forces were estimated by assuming a 25-MA plasma current decaying at 1 MA/ms while moving vertically. In addition to forces developed within the vessel, vertical loadings from the first wall/strong back assemblies and the divertor were applied to the vessel at their attachment points. The first 50 natural modes were also determined. The first mode's frequency was 6.0 Hz, and its shape is characterized by vertical displacement of the vessel inner leg. The predicted deformation of the vessel appeared similar to its first mode shape combined with radial contraction. Kinetic energy history from the analysis also correlated with the first mode frequency

  8. Observation of Dust in DIII-D Divertor and SOL by Visible Imaging

    International Nuclear Information System (INIS)

    Rudakov, D L; West, W P; Groth, M; Yu, J H; Wong, C C; Boedo, J A; Brooks, N H; Evans, T E; Fenstermacher, M E; Hollmann, E M; Hyatt, A W; Lasnier, C J; McLean, A G; Moyer, R A; Pigarov, A; Smirnov, R; Solomon, W M; Watkins, J G

    2007-01-01

    Dust is commonly found in fusion devices. Though generally of no concern in the present day machines, dust may pose serious safety and operational concerns for ITER. Micron-size dust usually dominates the samples collected from tokamaks. During a plasma discharge micron-size dust particles can become highly mobile and travel over distances of a few meters. Once inside the plasma, dust particles heat up to over 3000 K and emit thermal radiation that can be detected by visible imaging techniques. Observations of naturally occurring and artificially introduced dusts have been performed in DIII-D divertor and scrape-off layer (SOL) using standard frame rate CMOS cameras, a gated-intensified CID camera, and a fast-framing CMOS camera. In the first 2-3 plasma discharges after a vent with personnel entry inside the vacuum vessel ('dirty vent') dust levels were quite high with thousands of particles observed in each discharge. Individual particles moving at velocities of up to a few hundred m/s and breakup of larger particles into pieces were observed. After about 15 discharges dust was virtually gone during the stationary portion of a discharge, and appeared at much reduced levels during the plasma initiation and termination phases. After a few days of plasma operations (about 70 discharges) dust levels were further reduced to just a few observed events per discharge except in discharges with current disruptions that produced significant amounts of dust. An injection of a few milligram of micron-size (6 micron median diameter) carbon dust into a high-power lower single-null ELMing H-mode discharge with strike points swept across the lower divertor floor was performed. A significant increase of the core carbon radiation was observed for about 250 ms after the injection, as the total radiated power increased twofold. Dust particles from the injection were observed by the fast framing camera in the outboard SOL near the midplane. The amount of dust observed by the fast

  9. Particle control in DIII-D with helium glow discharge conditioning

    International Nuclear Information System (INIS)

    Jackson, G.L.; Taylor, T.S.; Taylor, P.L.

    1990-01-01

    Helium glow discharge conditioning of DIII-D is routinely used before every tokamak discharge to desorb hydrogen from the graphite tiles, which are the plasma facing surfaces for the floor, inner wall and top of the vessel. In addition to reducing hydrogen fuelling of the plasma by the graphite surfaces, helium glow discharges are also effective in removing low-Z impurities, primarily in the form of carbon monoxide and hydrocarbons, and this has permitted higher current divertor operation and more rapid recovery from tokamak disruptions. Since the implementation of repetitive helium glow wall conditioning, the parameter space in which tokamak discharges in DIII-D can be obtained has been expanded to include the first observations of limiter H-mode confinement, the Ohmic H-mode with periods of up to 150 ms that are free of edge localized modes, more reliable low q operation with volume averaged beta of up to 9.3%, improved control over locked modes and plasma discharges at lower electron density. (author). 37 refs, 12 figs, 1 tab

  10. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.; Fenstermacher, M.E.; Hill, D.N.; Hyatt, A.W.; Knoll, D.; Lasnier, C.J.; Lazarus, E.A.; Leonard, A.W.; Lippmann, S.I.; Mahdavi, M.A.; Maingi, R.; Meyer, W.; Moyer, R.A.; Petrie, T.W.; Porter, G.D.; Rensink, M.E.; Rognlien, T.D.; Schaffer, M.J.; Smith, J.P.; Staebler, G.M.; Stambaugh, R.D.; West, W.P.; Wood, R.D.

    1995-01-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while τ E remains similar 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ∼0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.))

  11. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, P.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.K.; Owen, L.; Matthews, G.

    1990-01-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p T < or approx.10.7%, P(auxiliary)< or approx.20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions. (orig.)

  12. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, T.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.; Owen, L.; Matthews, G.

    1990-06-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p approx-lt 3 MA, β T approx-lt 10.7%, P(auxiliary) approx-lt 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma surface interactions. 41 refs., 8 figs., 1 tab

  13. PHYSICS OF ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

    International Nuclear Information System (INIS)

    PETTY, C.C.; PRATER, R.; LUCE, T.C.; ELLIS, R.A.; HARVEY, R.W.; KINSEY, J.E.; LAO, L.L.; LOHR, J.; MAKOWSKI, M.A.

    2002-01-01

    OAK A271 PHYSICS OF ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The measured ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co and counter injection. The measured ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (Toray-GA) as well as a quasilinear Fokker-Planck model (CQL3D). The experimental ECCD is found to be in better agreement with the more complete Fokker-Planck calculation, especially for cases of high rf power density and/or loop voltage

  14. Divertor particle exhaust and wall inventory on DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Jackson, G.L.; Mahdavi, M.A.; Schaffer, M.J.; Wade, M.R.; Mioduszewski, P.K.; Hogan, J.T.; Klepper, C.C.; Haas, G.

    1995-01-01

    Many tokamaks achieve optimum plasma performance by achieving low recycling; various wall conditioning techniques including helium glow discharge cleaning (HeGDC) are routinely applied to help achieve low recycling. Many of these techniques allow strong, transient wall pumping, but they may not be effective for long-pulse tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), the Tokamak Physics Experiment (TPX), Tore Supra Continu, and JT-60SU. Continuous particle exhaust using an in-situ pumping scheme may be effective for wall inventory control in such devices. Recent particle balance experiments on the Tore Supra and DIII-D tokamaks demonstrated that the wall particle inventory could be reduced during a given discharge by use of continuous particle exhaust. In this paper the authors report the first results of wall inventory control and good performance with the in-situ DIII-D cryopump, replacing the HeGDC normally applied between discharges

  15. JT-60SA vacuum vessel manufacturing and assembly

    Energy Technology Data Exchange (ETDEWEB)

    Masaki, Kei, E-mail: masaki.kei@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Shibama, Yusuke K.; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The design of the JT-60SA vacuum vessel body was completed with the demonstration of manufacturing procedure by the mock-up fabrication of the 20 Degree-Sign upper half of VV. Black-Right-Pointing-Pointer The actual VV manufacturing has started since November 2009. Black-Right-Pointing-Pointer The first product of the VV 40 Degree-Sign sector was completed in May 2011. Black-Right-Pointing-Pointer A basic VV assembly scenario and procedure were studied to complete the 360 Degree-Sign VV including positioning method and joint welding. - Abstract: The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10 Degree-Sign segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content (<0.05%). The design temperatures of the VV at plasma operation and baking are 50 Degree-Sign C and 200 Degree-Sign C, respectively. In the double wall, boric-acid water is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas is circulated in the double wall after draining of the boric-acid water. The manufacturing of the VV started in November 2009 after a fundamental welding R and D and a trial manufacturing of 20 Degree-Sign upper half mock-up. The manufacturing of the first VV 40 Degree-Sign sector was completed in May 2011. A basic concept and required jigs of the VV assembly were studied. This paper describes the design and manufacturing of the vacuum vessel. A plan of VV assembly in torus hall is also presented.

  16. Demonstration of ITER operational scenarios on DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; DeBoo, J.C.; Ferron, J.R.; Jackson, G.L.; Luce, T.C.; Osborne, T.H.; Politzer, P.A.; Groebner, R.J.; Hyatt, A.W.; La Haye, R.J.; Petrie, T.W.; Petty, C.C.; Murakami, M.; Park, J.-M.; Reimerdes, H.; Budny, R.V.; Casper, T.A.; Holcomb, C.T.; Challis, C.D.; McKee, G.R.

    2010-01-01

    The DIII-D programme has recently initiated an effort to provide suitably scaled experimental evaluations of four primary ITER operational scenarios. New and unique features of this work are that the plasmas incorporate essential features of the ITER scenarios and anticipated operating characteristics; e.g. the plasma cross-section, aspect ratio and value of I/aB of the DIII-D discharges match the ITER design, with size reduced by a factor of 3.7. Key aspects of all four scenarios, such as target values for β N and H 98 , have been replicated successfully on DIII-D, providing an improved and unified physics basis for transport and stability modelling, as well as for performance extrapolation to ITER. In all four scenarios, normalized performance equals or closely approaches that required to realize the physics and technology goals of ITER, and projections of the DIII-D discharges are consistent with ITER achieving its goals of ≥400 MW of fusion power production and Q ≥ 10. These studies also address many of the key physics issues related to the ITER design, including the L-H transition power threshold, the size of edge localized modes, pedestal parameter scaling, the impact of tearing modes on confinement and disruptivity, beta limits and the required capabilities of the plasma control system. An example of direct influence on the ITER design from this work is a modification of the physics requirements for the poloidal field coil set at 15 MA, based on observations that the inductance in the baseline scenario case evolves to a value that lies outside the original ITER specification.

  17. Demonstration of ITER Operational Scenarios on DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; Budny, R.V.; DeBoo, J.C.; Ferron, J.R.; Jackson, G.L.; Luce, T.C.; Murakami, M.; Osborne, T.H.; Park, J.; Politzer, P.A.; Reimerdes, H.; Casper, T.A.; Challis, C.D.; Groebner, R.J.; Holcomb, C.T.; Hyatt, A.W.; La Haye, R.J.; McKee, G.R.; Petrie, T.W.; Petty, C.C.; Rhodes, T.L.; Shafer, M.W.; Snyder, P.B.; Strait, E.J; Wade, M.R.; Wang, G.; West, W.P.; Zeng, L.

    2008-01-01

    The DIII-D program has recently initiated an effort to provide suitably scaled experimental evaluations of four primary ITER operational scenarios. New and unique features of this work are that the plasmas incorporate essential features of the ITER scenarios and anticipated operating characteristics; e.g., the plasma cross-section, aspect ratio and value of I/aB of the DIII-D discharges match the ITER design, with size reduced by a factor of 3.7. Key aspects of all four scenarios, such as target values for β N and H 98 , have been replicated successfully on DIII-D, providing an improved and unified physics basis for transport and stability modeling, as well as for performance extrapolation to ITER. In all four scenarios normalized performance equals or closely approaches that required to realize the physics and technology goals of ITER, and projections of the DIII-D discharges are consistent with ITER achieving its goals of (ge) 400 MW of fusion power production and Q (ge) 10. These studies also address many of the key physics issues related to the ITER design, including the L-H transition power threshold, the size of ELMs, pedestal parameter scaling, the impact of tearing modes on confinement and disruptivity, beta limits and the required capabilities of the plasma control system. An example of direct influence on the ITER design from this work is a modification of the specified operating range in internal inductance at 15 MA for the poloidal field coil set, based on observations that the measured inductance in the baseline scenario case lay outside the original ITER specification

  18. Dust Studies in DIII-D and TEXTOR

    International Nuclear Information System (INIS)

    Rudakov, D.L.; Litnovsky, A.; West, W.P.; Yu, J.H.; Boedo, J.A.; Bray, B.D.; Brezinsek, S.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Hollmann, E.M.; Huber, A.; Hyatt, A.W.; Krasheninnikov, S.I.; Lasnier, C.J.; Moyer, R.A.; Pigarov, A.Y.; Philipps, V.; Pospieszczyk, A.; Smirnov, R.D.; Sharpe, J.P.; Solomon, W.M.; Watkins, J.G.; Wong, C.C.

    2009-01-01

    Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Direct heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust. Large flakes or debris falling into the plasma may result in a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micron-size dust in plasma discharges. In DIII-D, a sample holder filled with ∼30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure (∼0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. At the given configuration of the launch, the dust did not penetrate the core plasma and only moderately perturbed the edge plasma, as evidenced by an increase of the edge carbon content.

  19. Magnetic and electrical properties of ITER vacuum vessel steels

    International Nuclear Information System (INIS)

    Mergia, K.; Apostolopoulos, G.; Gjoka, M.; Niarchos, D.

    2007-01-01

    Full text of publication follows: Ferritic steel AISI 430 is a candidate material for the lTER vacuum vessel which will be used to limit the ripple in the toroidal magnetic field. The magnetic and electrical properties and their temperature dependence in the temperature range 300 - 900 K of AISI 430 ferritic stainless steels are presented. The temperature variation of the coercive field, remanence and saturation magnetization as well as electrical resistivity and the effect of annealing on these properties is discussed. (authors)

  20. Vacuum vessel port structures for ITER-FEAT

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th.

    2001-01-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed

  1. Vacuum vessel port structures for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th

    2001-11-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed.

  2. Upgrade of the DIII-D RF systems

    International Nuclear Information System (INIS)

    Callis, R.W.; Cary, W.P.; O'Neill, R.C.

    1995-10-01

    The DIII-D Advanced Tokamak Program requires the ability to modify the current density profile for extended time periods in order to achieve the improved plasma conditions now achieved with transient means. To support this requirement DIII-D has just completed a major addition to its ion cyclotron range of frequency (ICRF) systems. This upgrade project added two new fast wave current drive (FWCD) systems, with each system consisting of a 2 MW, 30 to 120 MHz transmitter, an all ceramic insulated transmission line, and water-cooled four-strap antenna. With this addition of 4 MW of FWCD power to the original 2 MW, 30 to 60 MHz capability, experiments can be performed with centrally localized current drive enhancement. For off-axis current modification, plans are in place to add 110 GHz electron cyclotron heating (ECH) power to DIII-D. Initially, 3 MW of power will be available with plans to increase the power to 6 MW and to 10 MW

  3. Using AORSA to simulate helicon waves in DIII-D

    International Nuclear Information System (INIS)

    Lau, C.; Blazevski, D.; Green, D. L.; Murakami, M.; Park, J. M.; Jaeger, E. F.; Berry, L. A.; Bertelli, N.; Pinsker, R. I.; Prater, R.

    2015-01-01

    Recent efforts have shown that helicon waves (fast waves at > 20ω ci ) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects

  4. Engineering design of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Bozek, A.S.

    1995-10-01

    A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor

  5. Overview of the DIII-D program computer systems

    International Nuclear Information System (INIS)

    McHarg, B.B. Jr.

    1997-11-01

    Computer systems pervade every aspect of the DIII-D National Fusion Research program. This includes real-time systems acquiring experimental data from data acquisition hardware; cpu server systems performing short term and long term data analysis; desktop activities such as word processing, spreadsheets, and scientific paper publication; and systems providing mechanisms for remote collaboration. The DIII-D network ties all of these systems together and connects to the ESNET wide area network. This paper will give an overview of these systems, including their purposes and functionality and how they connect to other systems. Computer systems include seven different types of UNIX systems (HP-UX, REALIX, SunOS, Solaris, Digital UNIX, Ultrix, and IRIX), OpenVMS systems (both BAX and Alpha), MACintosh, Windows 95, and more recently Windows NT systems. Most of the network internally is ethernet with some use of FDDI. A T3 link connects to ESNET and thus to the Internet. Recent upgrades to the network have notably improved its efficiency, but the demand for bandwidth is ever increasing. By means of software and mechanisms still in development, computer systems at remote sites are playing an increasing role both in accessing and analyzing data and even participating in certain controlling aspects for the experiment. The advent of audio/video over the interest is now presenting a new means for remote sites to participate in the DIII-D program

  6. Using AORSA to simulate helicon waves in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Lau, C., E-mail: lauch@ornl.gov; Blazevski, D.; Green, D. L.; Murakami, M.; Park, J. M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN (United States); Jaeger, E. F.; Berry, L. A. [XCEL Engineering, Inc., 1066 Commerce Park Dr., Oak Ridge, TN (United States); Bertelli, N. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Pinsker, R. I.; Prater, R. [General Atomics, San Diego, CA (United States)

    2015-12-10

    Recent efforts have shown that helicon waves (fast waves at > 20ω{sub ci}) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.

  7. Interprocess communication within the DIII-D plasma control system

    International Nuclear Information System (INIS)

    Piglowski, D.A.; Penaflor, B.G.; Ferron, J.R.

    1999-06-01

    The DIII-D tokamak fusion research experiment's real-time digital plasma control system (PCS) is a complex and ever evolving system. During a plasma experiment, it is tasked with some of the most crucial functions at DIII-D. Key responsibilities of the PCS involve sub-system control, data acquisition/storage, and user interface. To accomplish these functions, the PCS is broken down into individual components (both software and hardware), each capable of handling a specific duty set. Constant interaction between these components is necessary prior, during and after a standard plasma cycle. Complicating the matter even more is that some components, mostly those which deal with user interaction, may exist remotely, that is to say they are not part of the immediate hardware which makes up the bulk of the PCS. The four main objectives of this paper are to (1) present a brief outline of the PCS hardware/software and how they relate to each other; (2) present a brief overview of a standard DIII-D plasma cycle (a shot); (3) using three sets of PCS sub-systems, describe in more detail the communication processes; and (4) evaluate the benefits and drawbacks of said systems

  8. Design and analysis of the DIII-D radiative divertor water-cooled structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Phelps, R.D.; Redler, K.M.; Reis, E.E.

    1995-01-01

    The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electro-magnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

  9. Demonstration tests for manufacturing the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Onozuka, Masanori; Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro; Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi

    2007-01-01

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV

  10. Demonstration tests for manufacturing the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: katsusuke_shimizu@mhi.co.jp; Onozuka, Masanori [Mitsubishi Heavy Industries, Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan); Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)

    2007-10-15

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV.

  11. Development of a master model concept for DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy)

    2016-11-15

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  12. Development of a master model concept for DEMO vacuum vessel

    International Nuclear Information System (INIS)

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea; Bachmann, Christian; Di Gironimo, Giuseppe

    2016-01-01

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  13. Collaboration on DIII-D Five Year Plan

    International Nuclear Information System (INIS)

    Allen, S

    2003-01-01

    This document summarizes Lawrence Livermore National Laboratory's (LLNL) plan for fusion research on the DIII-D Tokamak, located at General Atomics (GA) in San Diego, California, in the time period FY04-FY08. This document is a companion document to the DIII-D Five-Year Program Plan; which hereafter will be referred to as the ''D3DPP''. The LLNL Collaboration on DIII-D is a task-driven program in which we bring to bear the full range of expertise needed to complete specific goals of plasma science research on the DIII-D facility. This document specifies our plasma performance and physics understanding goals and gives detailed plans to achieve those goals in terms of experimental leadership, code development and analysis, and diagnostic development. Our program is designed to be consistent with the long-term mission of the DIII-D program as documented in the D3DPP. The overall DIII-D Program mission is ''to establish the scientific basis for the optimization of the tokamak approach to fusion energy production''. LLNL Magnetic Fusion Energy (MFE) supports this mission, and we contribute to two areas of the DIII-D program: divertor physics and advanced tokamak (AT) physics. We lead or contribute to the whole cycle of research: experimental planning, diagnostic development, execution of experiments, and detailed analysis. We plan to continue this style in the next five years. DIII-D has identified three major research themes: AT physics, confinement physics, and mass transport. The LLNL program is part of the AT theme: measurement of the plasma current profile, and the mass transport theme: measurement and modeling of plasma flow. In the AT area, we have focused on the measurement and modeling of the current profile in Advanced Tokamak plasmas. The current profile, and it's effect on MHD stability of the high-β ''AT'' plasma are at the heart of the DIII-D program. LLNL has played a key role in the development of the Motional Stark Effect (MSE) diagnostic. Starting

  14. MFTF-α+T end cell vacuum vessel and nuclear shield trade studies

    International Nuclear Information System (INIS)

    Kirchner, J.

    1984-01-01

    Three separate and distinct vacuum vessel and nuclear shield trade studies were performed in series. The studies are: vacuum topology, nuclear shield location and composition, and water bulk shield location and material selection

  15. Experiments to Measure Hydrogen Release from Graphite Walls During Disruptions in DIII-D

    International Nuclear Information System (INIS)

    Hollmann, E.M.; Pablant, N.A.; Rudakov, D.L.; Boedo, J.A.; Brooks, N.H.; Jernigan, Thomas C.; Pigarov, A.Y.

    2009-01-01

    Spectroscopy and wall the bake-out measurements are performed in the DIII-D tokamak to estimate the amount of hydrogen stored in and released from the walls during disruptions. Both naturally occurring disruptions and disruptions induced by massive gas injection (MGI) are investigated. The measurements indicate that both types of disruptions cause a net release of order 10(21) hydrogen (or deuterium) atoms from the graphite walls. This is comparable to the pre-disruptions plasma particle inventory, so the released hydrogen is important for accurate modeling of disruptions. However, the amount of hydrogen released is small compared to the total saturated wall inventory of order 10(22)-10(23), So it appears that many disruptions are necessary to provide full pump-out of the vessel walls. (C) 2009 Published by Elsevier B.V.

  16. Particle exhaust schemes in the DIII-D advanced divertor configuration

    International Nuclear Information System (INIS)

    Menon, M.M.; Mioduszewski, P.K.

    1989-01-01

    For density control in long-pulse operation, the open divertor on the DIII-D tokamak will be equipped with a baffled chamber and a pumping system. The throat of the baffle chamber is sized to provide optimal pumping for the typical plasma equilibrium configuration. Severe limitations on the toroidal conductance of this baffle chamber require the use of in-vessel pumping to achieve the desired particle exhaust of about 25 Torr·l/s. Two separate pumping schemes are considered: an array of titanium getter modules based on the design developed by the Tore Supra team and a cryocondensation pump. The merits and demerits of each scheme are analyzed, and the design considerations introduced by the tokamak environment are brought out. 3 refs., 5 figs

  17. Development of a Closed Loop Simulator for Poloidal Field Control in DIII-D

    International Nuclear Information System (INIS)

    J.A. Leuer; M.L. Walker; D.A. Humphreys; J.R. Ferron; A. Nerem; B.G. Penaflor

    1999-01-01

    The design of a model-based simulator of the DIII-D poloidal field system is presented. The simulator is automatically configured to match a particular DIII-D discharge circuit. The simulator can be run in a data input mode, in which prior acquired DIII-D shot data is input to the simulator, or in a stand-alone predictive mode, in which the model operates in closed loop with the plasma control system. The simulator is used to design and validate a multi-input-multi-output controller which has been implemented on DIII-D to control plasma shape. Preliminary experimental controller results are presented

  18. Clearance potential of ITER vacuum vessel activated materials

    International Nuclear Information System (INIS)

    Cepraga, D.G.; Cambi, G.; Frisoni, M.

    2002-01-01

    To demonstrate fusion's environmental attractiveness over the entire life cycle, a waste analysis is mandatory. The clearance is recommended by IAEA for releasing activated solid materials from regulatory control and for waste management policy. The paper focuses on the approach used to support waste analyses for ITER Generic Site Safety Report. The Material Unconditional Clearance Index of all the materials/zones on the equatorial mid-plane of ITER machine have been evaluated, based on IAEA-TECDOC-855. The Bonami-Nitawl-XSDNRPM sequence of the Scale-4.4a code system (using Vitenea-J library) has been firstly used for radiation transport analyses. Then the Anita-2000 code package is used for the activation calculation. The paper presents also, as an example, an application of the clearance indexes estimation for the ITER vacuum vessel materials. The results of the Anita-2000 have been compared with those obtained using the Fispact-99 activation code. (author)

  19. Manufacturing technology development for vacuum vessel and plasma facing components

    International Nuclear Information System (INIS)

    Laitinen, Arttu; Liimatainen, Jari; Hallila, Pentti

    2005-01-01

    Vacuum vessel and plasma facing components of the ITER construction including shield modules and primary first wall panels have great impact on the production costs and reliability of the installation. From the manufacturing technology point of view, accuracy of shape, properties of the various austenitic stainless steel/austenitic stainless steel interfaces or CuCrZr/austenitic stainless steel interfaces as well as those of the base materials are crucial for technical reliability of the construction. The current approach in plasma facing components has been utilisation of solid-HIP technology and solid-powder-HIP technology. Due to the large size of especially shield modules shape, control of the internal cavities and cooling channels is extremely demanding. This requires strict control of the raw materials and manufacturing parameters

  20. Hydrogen/hydrocarbon explosions in the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Goranson, P.L.

    1992-01-01

    The consequences of H 2 /hydrocarbon detonations in the vacuum vessel (torus) of the International Thermonuclear Experimental Reactor (ITER) have been studied. The most likely scenario for such a detonation involves a water leak into the torus and a vent of the torus to atmosphere, permitting the formation of an explosive fuel-air mixture. The generation of fuel gases and possible sources of air or oxygen are reviewed, and the severity and effects of specific fuel-air mixture explosions are evaluated. Detonation or deflagration of an explosive mixture could result in pressures exceeding the maximum allowable torus pressure. Further studies to examine the design details and develop an event-tree study of events following a gas detonation are recommended

  1. Assessment of the dynamic behaviours of the ITER Vacuum Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Blocki, J., E-mail: jacek.blocki@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Combescure, D. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Mazzone, G. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► The cyclic symmetry structure with special boundary conditions has been analyzed. ► Results based on the FE solid model and on the FE shell model have been compared. ► The effect of the missing mass contained has been checked. -- Abstract: The dynamic behaviour of the ITER Vacuum Vessel (VV) under seismic loads will be assessed by carrying out the modal analysis and then by applying the response spectrum method which describes earthquake motions. The effect of the missing mass is included in this last analysis. Numerical results are based on two different Finite Element (FE) models and on three different methods by which natural frequencies and mode shapes are defined. It means, it is applied the cyclic symmetry method, the component mode synthesis method and the 360° FE model of the VV. Comparisons between obtained results for the different models and methods are presented.

  2. Design study of a new vacuum vessel for Doublet III

    International Nuclear Information System (INIS)

    Rawls, J.M.; Davis, L.G.; Anderson, P.M.

    1980-10-01

    The principal thrust of the project was to examine a single design in enough depth to gain confidence in the feasibility and desirability of specific design features. However, a valuable spin-off of the project was to develop information of a more generic character to aid in future studies of possibilities for Doublet III. For example, we now feel that Doublet III can be reconfigured with any of a variety of new vacuum vessels, poloidal coil sets, and auxiliary heating systems within three years of project initiation, a period that is short compared to the time scale for developing a completely new facility. In addition, this can be accomplished at a fraction of the cost required to develop a comparable facility

  3. Design and R and D for the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R.; Koizumi, K.; Kuzmin, E.; Maisonnier, D.; Nelson, B.

    1998-01-01

    The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

  4. Design and R and D for the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R. [ITER Joint Work Site, Garching (Germany); Koizumi, K. [Japan Atomic Energy Research Inst., Naka (Japan); Kuzmin, E. [Efremov Insitute, Saint Petersburg (Russian Federation); Maisonnier, D. [NET Team, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

  5. IWR-solution for the ITER vacuum vessel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Wu, H., E-mail: huapeng@lut.fi [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Handroos, H. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Pela, P. [Tekes (Finland); Wang, Y. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland)

    2011-10-15

    The assembly of ITER vacuum vessel (VV) is still a very big challenge as the process can only be done from inside the VV. The welding of the VV assembly is carried out using the dedicated robotic systems. The main functions of the robots are: (i) measuring the actual space between every two sectors, (ii) positioning of the 150 kg splice plates between the sector shells, (iii) welding the splice plates to the sector shells, (iv) NDT of the welds, (v) repairing, including machining of the welds, (vi) He-leak tests of the welds, and (vii) the non-planned functions that may turn out. This paper presents a reasonable method to assemble the ITER VV. In this article, one parallel mobile robot, running on the track rail fixed on the wall inside the VV, is designed and tested. The assembling process, carried out by the mobile robot together with the welding robot, is presented.

  6. Neutral beam current drive scaling in DIII-D

    International Nuclear Information System (INIS)

    Porter, G.D.; Bhadra, D.K.; Burrell, K.H.

    1989-03-01

    Neutral beam current drive scaling experiments have been carried out on the DIII-D tokamak at General Atomics. These experiments were performed using up to 10 MW of 80 keV hydrogen beams. Previous current drive experiments on DIII-D have demonstrated beam driven currents up to 340 kA. In the experiments reported here we achieved beam driven currents of at least 500 kA, and have obtained operation with record values of poloidal beta (εβ/sub p/ = 1.4). The beam driven current reported here is obtained from the total plasma current by subtracting an estimate of the residual Ohmic current determined from the measured loop voltage. In this report we discuss the scaling of the current drive efficiency with plasma conditions. Using hydrogen neutral beams, we find the current drive efficiency is similar in Deuterium and Helium target plasmas. Experiments have been performed with plasma electron temperatures up to T/sub e/ = 3 keV, and densities in the range 2 /times/ 10 19 m/sup /minus/3/ 19 m/sup /minus/3/. The current drive efficiency (nIR/P) is observed to scale linearly with the energy confinement time on DIII-D to a maximum of 0.05 /times/ 10 20 m/sup /minus/2/ A/W. The measured efficiency is consistent with a 0-D theoretical model. In addition to comparison with this simple model, detailed analysis of several shots using the time dependent transport code ONETWO is discussed. This analysis indicates that bootstrap current contributes approximately 10--20% of the the total current. Our estimates of this effect are somewhat uncertain due to limited measurements of the radial profile of the density and temperatures. 4 refs., 1 fig., 1 tab

  7. Advances in Integrated Plasma Control on DIII-D

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Humphreys, D.A.

    2006-01-01

    The DIII-D experimental program in advanced tokamak (AT) physics requires extremely high performance from the DIII-D plasma control system (PCS) [B.G.Penaflor, et al., 4 th IAEA Tech. Mtg on Control and Data Acq., San Diego, CA (2003)], including simultaneous and highly accurate regulation of plasma shape, stored energy, density, and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of several new model-based plasma controllers on DIII-D. We discuss experimental use of advanced shape control algorithms containing nonlinear techniques for improving control of steady state plasmas, model-based controllers for optimal rejection of edge localized mode disturbances during resistive wall mode stabilization, model-based controllers for neoclassical tearing mode stabilization, including methods for maximizing stabilization effectiveness with substantial constraints on available power, model-based integrated control of plasma rotation and beta, and initial experience in development of model-based controllers for advanced tokamak current profile modification. The experience gained from DIII-D has been applied to the development of control systems for the EAST and KSTAR tokamaks. We describe the development of the control software, hardware, and model-based control algorithms for these superconducting tokamaks, with emphasis on relevance of

  8. Plasma rotation and rf heating in DIII-D

    International Nuclear Information System (INIS)

    DeGrassie, J.S.; Baker, D.R.; Burrell, K.H.

    1999-05-01

    In a variety of discharge conditions on DIII-D it is observed that rf electron heating reduces the toroidal rotation speed and core ion temperature. The rf heating can be with either fast wave or electron cyclotron heating and this effect is insensitive to the details of the launched toroidal wavenumber spectrum. To date all target discharges have rotation first established with co-directed neutral beam injection. A possible cause is enhanced ion momentum and thermal diffusivity due to electron heating effectively creating greater anomalous viscosity. Another is that a counter directed toroidal force is applied to the bulk plasma via rf driven radial current

  9. Experiments at high elongations in DIII-D

    International Nuclear Information System (INIS)

    Lazarus, E.A.; Turnbull, A.D.; Kellman, A.G.; Ferron, J.R.; Helton, F.J.; Lao, L.L.; Leuer, J.A.; Strait, E.J.; Taylor, T.S.

    1990-06-01

    In this paper we discuss the limitation to elongation observed in D-shaped plasmas in the DIII-D tokamak. We find that as the triangularity is increased and ell i is decreased that the n = 0 mode takes on an increasingly non-rigid character. Our analysis shows two aspects of the behavior; first, an increasing variation of the m/n = 1/0 component across flux surfaces and second, an increase in the relative amplitude of a m/n = 3/0 component which couples to the m/n = 1/0 component and further destabilizes the mode

  10. Performance characteristics of the DIII-D advanced divertor cryopump

    International Nuclear Information System (INIS)

    Menon, M.M.; Maingi, R.; Wade, M.R.; Baxi, C.B.; Campbell, G.L.; Holtrop, K.L.; Hyatt, A.W.; Laughon, G.J.; Makariou, C.C.; Mahdavi, M.A.; Reis, E.E.; Schaffer, M.J.; Schaubel, K.M.; Scoville, J.T.; Smith, J.P.; Stambaugh, R.D.

    1993-10-01

    A cryocondensation pump, cooled by forced flow of two-phase helium, has been installed for particle exhaust from the divertor region of the DIII-D tokamak. The Inconel pumping surface is of coaxial geometry, 25.4 mm in outer diameter and 11.65 m in length. Because of the tokamak environment, the pump is designed to perform under relatively high pulsed heat loads (300 Wm -2 ). Results of measurements made on the pumping characteristics for D 2 , H 2 , and Ar are discussed

  11. Quiescent double barrier regime in the DIII-D tokamak.

    Science.gov (United States)

    Greenfield, C M; Burrell, K H; DeBoo, J C; Doyle, E J; Stallard, B W; Synakowski, E J; Fenzi, C; Gohil, P; Groebner, R J; Lao, L L; Makowski, M A; McKee, G R; Moyer, R A; Rettig, C L; Rhodes, T L; Pinsker, R I; Staebler, G M; West, W P

    2001-05-14

    A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the E x B shearing rate.

  12. Plasma rotation and rf heating in DIII-D

    International Nuclear Information System (INIS)

    Grassie, J. S. de; Baker, D. R.; Burrell, K. H.; Greenfield, C. M.; Lin-Liu, Y. R.; Luce, T. C.; Petty, C. C.; Prater, R.; Heidbrink, W. W.; Rice, B. W.

    1999-01-01

    In a variety of discharge conditions on DIII-D it is observed that rf electron heating reduces the toroidal rotation speed and core ion temperature. The rf heating can be with either fast wave or electron cyclotron heating and this effect is insensitive to the details of the launched toroidal wavenumber spectrum. To date all target discharges have rotation first established with co-directed neutral beam injection. A possible cause is enhanced ion momentum and thermal diffusivity due to electron heating effectively creating greater anomalous viscosity. Another is that a counter directed toroidal force is applied to the bulk plasma via rf driven radial current. (c) 1999 American Institute of Physics

  13. Ion Bernstein wave antenna design for DIII-D

    International Nuclear Information System (INIS)

    Phelps, R.D.; Mayberry, M.J.; Pinsker, R.J.

    1989-01-01

    An array of two toroidal loop antennas has been designd and installed on the DIII-D tokamak to carry out Ion Bernstein Wave (IBW) heating experiments. The antenna will operate at the 2 MW level and provide direct excitation of the IBW over the frequency range of 30-60 MHz. This device will permit the study of coupling th IBW to divertor plasmas and will provide a menas for improving the confinement and stability of high beta plasmas through localized off-axis heating. This paper describes both the mechanical and electromagnetic design of the IBW antenna. (author). 2 refs.; 4 figs.; 1 tab

  14. Thermal deposition analysis during disruptions on DIII-D using infrared scanners

    International Nuclear Information System (INIS)

    Lee, R.L.; Hyatt, A.W.; Kellman, A.G.; Taylor, P.L.; Lasnier, C.J.

    1995-12-01

    The DIII-D tokamak generates plasma discharges with currents up to 3 MA and auxiliary input power up to 20 MW from neutral beams and 4 MW from radio frequency systems. In a disruption, a rapid loss of the plasma current and internal thermal energy occurs and the energy is deposited onto the torus graphite wall. Quantifying the spatial and temporal characteristics of the heat deposition is important for engineering and physics-related issues, particularly for designing future machines such as ITER. Using infrared scanners with a time resolution of 120 micros, measurements of the heat deposition onto the all-graphite walls of DIII-D during two types of disruptions have been made. Each scanner contains a single point detector sensitive to 8--12 microm radiation, allowing surface temperatures from 20 C to 2,000 C to be measured. A zinc selenide window that transmits in the infrared is used as the vacuum window. Views of the upper and lower divertor regions and the centerpost provide good coverage of the first wall for single and double null divertor discharges. During disruptions, the thermal energy is not deposited evenly onto the inner surface of the tokamak, but is deposited primarily in the divertor region when operating diverted discharges. Analysis of the heat deposition during a radiative collapse disruption of a 1.5 MA discharge revealed power densities of 300--350 MW/m 2 in the divertor region. During the thermal quench of the disruption, the energy deposited onto the divertor region was more than 70% of the stored thermal energy in the discharge prior to the disruption. The spatial distribution and temporal behavior of power deposition during high β disruptions will also be presented

  15. The 110 GHz Gyrotron System on DIII-D: Gyrotron Tests and Physics Results

    International Nuclear Information System (INIS)

    Lohr, J.; Calahan, P.; Callis, R.W.

    1999-01-01

    The DIII-D tokamak has installed a system with three gyrotrons at the 1 MW level operating at 110 GHz. Physics experiments on electron cyclotron current drive, heating, and transport have been performed. Good efficiency has been achieved both for on-axis and off-axis current drive with relevance for control of the current density profile leading to advanced regimes of tokamak operation, although there is a difference between off-axis ECCD efficiency inside and outside the magnetic axis. Heating efficiency is excellent and electron temperatures up to 10 keV have been achieved. The gyrotron system is versatile, with poloidal scan and control of the polarization of the injected rf beam. Phase correcting mirrors form a Gaussian beam and focus it into the waveguide. Both perpendicular and oblique launch into the tokamak have been used. Three different gyrotron designs are installed and therefore unique problems specific to each have been encountered, including parasitic oscillations, mode hops during modulation and polarization control problems. Two of the gyrotrons suffered damage during operations, one due to filament failure and one due to a vacuum leak. The repairs and subsequent testing will be described. The transmission system uses evacuated, windowless waveguide and the three gyrotrons have output windows of three different materials. One gyrotron uses a diamond window and generates a Gaussian beam directly. The development of the system and specific tests and results from each of the gyrotrons will be presented. The DIII-D project has committed to an upgrade of the system, which will add three gyrotrons in the 1 MW class, all using diamond output windows, to permit operation at up to ten seconds per pulse at one megawatt output for each gyrotron

  16. Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel

    Science.gov (United States)

    Maheshwari, A.; Pathak, H. A.; Mehta, B. K.; Phull, G. S.; Laad, R.; Shaikh, M. S.; George, S.; Joshi, K.; Khan, Z.

    2017-04-01

    ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got

  17. Utilization of vanadium alloys in the DIII-D Radiative Divertor Program

    International Nuclear Information System (INIS)

    Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

    1995-10-01

    Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics (GA), in conjunction with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan for the utilization of vanadium alloys as part of the Radiative Divertor (RD) upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy (DOE). This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components, and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming Radiative Divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development (R and D) efforts to support fabrication development and to resolve key issues related to environmental effects

  18. Utilization of vanadium alloys in the DIII-D radiative divertor program

    International Nuclear Information System (INIS)

    Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

    1996-01-01

    Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics in conjunction with Argonne National Laboratory and Oak Ridge National Laboratory has developed a plan for the utilization of vanadium alloys as part of the radiative divertor upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy. This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming radiative divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development efforts to support fabrication development and to resolve key issues related to environmental effects. (orig.)

  19. Lawrence Livermore National Laboratory DIII-D cooperation: 1987 annual report

    International Nuclear Information System (INIS)

    Allen, S.L.; Calderon, M.O.; Ellis, R.M.

    1988-01-01

    This report summarizes the Lawrence Livermore National Laboratory (LLNL) DIII-D cooperation during FY87. The LLNL participation in DIII-D concentrated on three principal areas: ECH and current-drive physics, divertor and edge physics, and tokamak operations. These topics are dicussed in this report. 27 refs., 11 figs

  20. Dust processing device for inside of vacuum vessel of thermonuclear reactor

    International Nuclear Information System (INIS)

    Okumura, Atsushi; Tsujimura, Seiichi; Takahashi, Kenji; Ueda, Yasutoshi; Kuwata, Masayasu; Onozuka, Masaki.

    1995-01-01

    The device of the present invention can occasionally recover dusts in a vacuum vessel of a thermonuclear reactor. In addition, fine powdery dusts are never scattered to the vacuum vessel. Namely, a processing device main body comprises a locally sealed space in the vacuum vessel. A blow-up device blows up and floats dusts accumulated in the vacuum vessel to the processing device main body. A discharge plate electrically charges the floating dusts by discharge. An electrode collects the charged dusts. Collected dusts are recovered together with a pressurized gas through a dust recovering port to the outside of the processing device. With such a constitution, it is not necessary to release the vacuum vessel to the atmosphere and evacuate after the completion of the collection of the dusts on every time when the dusts are generated as in the prior art. It is no more necessary for an operator to enter into the vacuum vessel and recover the dusts. Since fine powdery dusts are never scattered in the vacuum vessel, no undesired effects are given to exhaustion facilities and instruments of the vacuum vessel. (I.S.)

  1. Dust processing device for inside of vacuum vessel of thermonuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Atsushi; Tsujimura, Seiichi; Takahashi, Kenji; Ueda, Yasutoshi; Kuwata, Masayasu; Onozuka, Masaki

    1995-05-02

    The device of the present invention can occasionally recover dusts in a vacuum vessel of a thermonuclear reactor. In addition, fine powdery dusts are never scattered to the vacuum vessel. Namely, a processing device main body comprises a locally sealed space in the vacuum vessel. A blow-up device blows up and floats dusts accumulated in the vacuum vessel to the processing device main body. A discharge plate electrically charges the floating dusts by discharge. An electrode collects the charged dusts. Collected dusts are recovered together with a pressurized gas through a dust recovering port to the outside of the processing device. With such a constitution, it is not necessary to release the vacuum vessel to the atmosphere and evacuate after the completion of the collection of the dusts on every time when the dusts are generated as in the prior art. It is no more necessary for an operator to enter into the vacuum vessel and recover the dusts. Since fine powdery dusts are never scattered in the vacuum vessel, no undesired effects are given to exhaustion facilities and instruments of the vacuum vessel. (I.S.).

  2. Automated Calculation of DIII-D Neutral Beam Availability

    International Nuclear Information System (INIS)

    Phillips, J.C.; Hong, R.M.; Scoville, B.G.

    1999-01-01

    The neutral beam systems for the DIII-D tokamak are an extremely reliable source of auxiliary plasma heating, capable of supplying up to 20 MW of injected power, from eight separate beam sources into each tokamak discharge. The high availability of these systems for tokamak operations is sustained by careful monitoring of performance and following up on failures. One of the metrics for this performance is the requested injected power profile as compared to the power profile delivered for a particular pulse. Calculating this was a relatively straightforward task, however innovations such as the ability to modulate the beams and more recently the ability to substitute an idle beam for one which has failed during a plasma discharge, have made the task very complex. For example, with this latest advance it is possible for one or more beams to have failed, yet the delivered power profile may appear perfect. Availability used to be manually calculated. This paper presents the methods and algorithms used to produce a system which performs the calculations based on information concerning the neutral beam and plasma current waveforms, along with post-discharge information from the Plasma Control System, which has the ability to issue commands for beams in real time. Plots representing both the requested and actual power profiles, along with statistics, are automatically displayed and updated each shot, on a web-based interface viewable both at DIII-D and by our remote collaborators using no-cost software

  3. A phase contrast interferometer on DIII-D

    International Nuclear Information System (INIS)

    Coda, S.; Porkolab, M.; Carlstrom, T.N.

    1992-04-01

    A novel imaging diagnostic has recently become operational on the DIII-D tokamak for the study of density fluctuations at the outer edge of the plasma. The phase contrast imaging approach overcomes the limitations of conventional scattering techniques in the spectral range of interest for transport-related phenomena, by allowing detection of long wavelength modes (up to 7.6 cm) with excellent spatial resolution (5 mm) in the radial direction. Additional motivation for the diagnostic is provided by wave-plasma interactions during heating and current drive experiments in the Ion Cyclotron range of frequencies. Density perturbations of 4 x 10 7 cm -3 with a 1 MHz bandwidth can be resolved. The diagnostic employs a 7.6 cm diameter CO 2 laser beam launched vertically across the plasma edge. An image of the plasma is then created on a 16-element detector array: the detector signals are directly proportional to the density fluctuations integrated along each chord. Wavelengths and correlation lengths can be inferred from the spatial mapping. The phase contrast method and its application to DIII-D are described and tests and first plasma data are presented

  4. Tritium in the DIII-D carbon tiles

    International Nuclear Information System (INIS)

    Taylor, P.L.; Kellman, A.G.; Lee, R.L.

    1993-06-01

    The amount of tritium in the carbon tiles used as a first wall in the DIII-D tokamak was measured recently when the tiles were removed and cleaned. The measurements were made as part of the task of developing the appropriate safety procedures for processing of the tiles. The surface tritium concentration on the carbon tiles was surveyed and the total tritium released from tile samples was measured in test bakes. The total tritium in all the carbon tiles at the time the tiles were removed for cleaning is estimated to be 15 mCi and the fraction of tritium retained in the tiles from DIII-D operations has a lower bound of 10%. The tritium was found to be concentrated in a narrow surface layer on the plasma facing side of the tile, was fully released when baked to 1,000 degree C, and was released in the form of tritiated gas (DT) as opposed to tritiated water (DTO) when baked

  5. Development of a radiative divertor for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Campbell, R.B. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Hill, D.N. [Lawrence Livermore National Lab., CA (United States); Hyatt, A.W. [General Atomics, San Diego, CA (United States); Knoll, D.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Lippmann, S.I. [General Atomics, San Diego, CA (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Maingi, R. [Oak Ridge National Lab., TN (United States); Meyer, W. [Lawrence Livermore National Lab., CA (United States); Moyer, R.A. [California Univ., Los Angeles, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Rensink, M.E. [Lawrence Livermore National Lab., CA (United States); Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States); Smith, J.P. [General Atomics, San Diego, CA (United States); Staebler, G.M. [General Atomics, San Diego, CA (United States); Stambaugh, R.D. [General Atomics, San Diego, CA (United States); West, W.P. [General Atomics, San Diego, CA (United States); Wood, R.D. [Lawrence Livermore National Lab., CA (United States)

    1995-04-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while {tau}{sub E} remains similar 2 times ITER-89P scaling. However, n{sub e} increases with D{sub 2} puffing, and Z{sub eff} increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity ({delta}{approx}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.)).

  6. The back transition and hysteresis effects in DIII-D

    International Nuclear Information System (INIS)

    Thomas, D.M.; Groebner, R.J.; Burrell, K.H.; Osborne, T.H.; Carlstrom, T.N.

    1997-09-01

    The back transition from H-mode to L-mode has been studied on DIII-D as a part of the investigation of the L-H transition power threshold scaling. Based on a density-dependent scaling for the H-mode power threshold, ITER will require substantial hysteresis in this parameter to remain in H-mode as n e rises. Defining the hysteresis in terms of the ratio of sustaining to threshold power, P HL /P LH may need to be as small as 50% for ITER. Operation of DIII-D at injection powers slightly above the H-mode threshold results in an oscillatory behavior with multiple forward-backward transitions in the course of a discharge. These discharges represent a unique system for studying various control parameters that may influence the H↔L state transition. Careful analysis of the power flow through the edge gives values for the sustaining power which are well below the corresponding threshold powers (P HL /P LH = 35--70%), indicating substantial hysteresis can be achieved in this parameter. Studies of other control parameter candidates such as edge temperature during the back transitions are less clear: the amount of hysteresis seen in these parameters, if any, is primarily dependent on the nature (ELMing, ELM-free) of the parent H-state

  7. UEDGE code comparisons with DIII-D bolometer data

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, J.M.

    1994-12-01

    This paper describes the work done to develop a bolometer post processor that converts volumetric radiated power values taken from a UEDGE solution, to a line integrated radiated power along chords of the bolometers in the DIII-D tokamak. The UEDGE code calculates plasma physics quantities, such as plasma density, radiated power, or electron temperature, and compares them to actual diagnostic measurements taken from the scrape off layer (SOL) and divertor regions of the DIII-D tokamak. Bolometers are devices measuring radiated power within the tokamak. The bolometer interceptors are made up of two complete arrays, an upper array with a vertical view and a lower array with a horizontal view, so that a two dimensional profile of the radiated power may be obtained. The bolometer post processor stores line integrated values taken from UEDGE solutions into a file in tabular format. Experimental data is then put into tabular form and placed in another file. Comparisons can be made between the UEDGE solutions and actual bolometer data. Analysis has been done to determine the accuracy of the plasma physics involved in producing UEDGE simulations.

  8. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  9. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  10. Remote collaboration and data access at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schissel, D.P.

    1998-09-01

    As the number of on-site and remote collaborators has increased, the demands on the DIII-D National Program's computational infrastructure has become more severe. The Director of the DIII-D Program recognized the increased importance of computers in carrying out the DIII-D mission and in late 1997 formed the Data Analysis Programming Group. Utilizing both software and hardware improvements, this new group has been charged with increasing the DIII-D data analysis throughput and data retrieval rate. Understanding the importance of the remote collaborators, this group has developed a long term plan that will allow for fast 24 hour data access (7x24) with complete documentation and a set of data viewing and analysis tools that can be run either on the collaborators' or DIII-D's computer systems. This paper presents the group's long term plan and progress to date

  11. Materials requirements for the ITER vacuum vessel and in-vessel components - approaching the construction phase

    International Nuclear Information System (INIS)

    Barabash, V.; Ioki, K.; Pick, M.; Girard, J.P.; Merola, M.

    2007-01-01

    Full text of publication follows: The ITER activities are fully devoted toward its construction. In accordance with the ITER integrated project schedule, the procurement specifications for the manufacturing of the Vacuum Vessel should be prepared by March 2008 and the procurement specifications for the in-vessel components (first wall/blanket, divertor) by 2009. To update the design, considering design and technology evolution, the ITER Design Review has been launched. Among the various topics being discussed are the important issues related to selection of materials, material procurement, and assessment of performance during operation. The main requirements related to materials for the vacuum vessel and the in-vessel components are summarized in the paper. The specific licensing requirements are to be followed for structural materials of pressure and nuclear pressure equipment components for construction of ITER. In addition, the procurements in ITER will be done mostly 'in-kind' and it is assumed that materials for these components will be produced by different Parties. However, in accordance with the regulatory requirements and quality requirements for operation, common specifications and the general rules to fulfill these requirements are to be adopted. For some ITER components (e.g. first wall, divertor high heat flux components), the ultimate qualification of the joining technologies (Be/Cu, SS/Cu, CFC/Cu, W/Cu) is under final evaluation. Successful accomplishment of the qualification program will allow to proceed with procurements of the components for ITER. The criteria for acceptance of these components and materials after manufacturing are described and the main results will be reported. Additional materials issues, which come from the on-going manufacturing R and D program, will be also described. Finally, further materials activity during the construction phase, needs for final qualification and acceptance of materials are discussed. (authors)

  12. Progress and Achievements on the R&D Activities for ITER Vacuum Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Research Institute (JAERI); Koizumi, K. [Japan Atomic Energy Research Institute (JAERI); Takahashi, H. [Japan Atomic Energy Research Institute (JAERI); Onozuka, M. [ITER Joint Central Team, Garching, Germany; Ioki, K. [ITER Joint Central Team, Garching, Germany; Kuzumin, E. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Krylov, V. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Maslakowski, J. [Oak Ridge National Laboratory (ORNL); Nelson, Brad E [ORNL; Jones, L. [Max-Planck Institute, Garching, Germany; Danner, W. [Max-Planck Institute, Garching, Germany; Maisonnier, D. [Max-Planck Institute, Garching, Germany

    2001-01-01

    The ITER vacuum vessel (VV) is designed to be large double-walled structure with a D-shaped crosssection. The achievable fabrication tolerance of this structure was unknown due to the size and complexity of shape. The Full-scale Sector Model of ITER Vacuum Vessel, which was 15m in height, was fabricated and tested to obtain the fabrication and assembly tolerances. The model was fabricated within the target tolerance of 5mm and welding deformation during assembly operation was obtained. The port structure was also connected using remotized welding tools to demonstrate the basic maintenance activity. In parallel, the tests of advanced welding, cutting and inspection system were performed to improve the efficiency of fabrication and maintenance of the Vacuum Vessel. These activities show the feasibility of ITER Vacuum Vessel as feasible in a realistic way. This paper describes the major progress, achievement and latest status of the R&D activities on the ITER vacuum vessel.

  13. Plasma discharge in ferritic first wall vacuum vessel of the Hitachi Tokamak HT-2

    International Nuclear Information System (INIS)

    Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio

    1997-01-01

    A tokamak discharge with ferritic material first wall was tried successfully. The Hitachi Tokamak HT-2 had a stainless steel SUS304 vacuum vessel and modified to have a ferritic plate first wall for experiments to investigate the possibility of ferritic material usage in magnetic fusion devices. The achieved vacuum pressure and times used for discharge cleaning was roughly identical with the stainless steel first wall or the original HT-2. We concluded that ferritic material vacuum vessel is possible for tokamaks. (author)

  14. Fabrication of a 1200 kg Ingot of V-4Cr-4Ti for the DIII-D Radiative Divertor Program

    International Nuclear Information System (INIS)

    Johnson, W.R.; Smith, J.P.

    1998-01-01

    Vanadium chromium titanium alloys are attractive materials for fusion reactors because of their high temperature capability and their potential for low neutron active and rapid activation decay. A V-4Cr-4Ti alloy has been selected in the U.S. as the current leading candidate vanadium alloy for future use in fusion reactor structural applications. General Atomics (GA), in conjunction with the Department of Energy's (DOE) DIII-D Program, is carrying out a plan for the utilization of this vanadium alloy in the DIII-D tokamak. The plan will culminate in the fabrication, installation, and operation of a V-4Ti alloy structure in the DIII-D Radiative Divertor (RD) upgrade. The deployment of vanadium alloy will provide a meaningful step in the development and technology acceptance of this advanced material for future fusion power devices. Under a GA contract and material specification, an industrial scale 1200 kg heat (ingot) of a V-4Cr-4Ti alloy has been produced and converted into product forms by Wah Chang of Albany, Oregon (WCA). To assure the proper control of minor and trace impurities which affect the mechanical and activation behavior of this vanadium alloy, selected lots of raw vanadium base metal were processed by aluminothermic reduction of high purity vanadium oxide, and were then electron beam melted into two high purity vanadium ingots. The ingots were then consolidated with high purity Cr and Ti, and double vacuum-arc melted to obtain a 1200 kg V-4Cr-4Ti alloy ingot. Several billets were extruded from the ingot, and were then fabricated into plate, sheet, and rod at WCA. Tubing was subsequently processed from plate material. The chemistry and fabrication procedures for the product forms were specified on the basis of experience and knowledge gained from DOE Fusion Materials Program studies on previous laboratory scale heats and a large scale ingot (500 kg)

  15. Engineering and design of a CO2 phase contrast interferometer system for DIII-D

    International Nuclear Information System (INIS)

    Phelps, R.D.; Coda, S.

    1994-11-01

    This report describes the development of a CO 2 laser interferometer system, the engineering, design and installation of the hardware, and the selection of materials specific to the requirements of a CO 2 laser diagnostic. A brief description of system operation is included. A phase contrast interferometer diagnostic has been designed and installed on the DIII-D tokamak to enhance studies of the physical characteristics of plasma turbulence, and specifically to analyze plasma density fluctuations in the boundary region of the plasma. A 20 watt CO 2 laser beam, operating at the 10.6 micron wavelength, is expanded to a diameter of 76 mm and directed through a series of mirrors which provide for entry of the beam into the vessel at a point 70 cm above the midplane at the 285 degree toroidal location. After being reflected from a mirror inside the vessel, the beam is directed downward so that it passes through the edge of the plasma immediately in front of a four-strap fast wave current drive rf antenna. The laser beam is then reflected by a second internal mirror and exits the vessel 70 cm below the midplane (also at 285 degrees) returning to an optical table through a final series of external steering mirrors

  16. Structural analysis of the ITER vacuum vessel from disruption loading with halo asymmetry

    International Nuclear Information System (INIS)

    Riemer, B.W.; Sayer, R.O.

    1996-01-01

    Static structural analyses of the ITER vacuum vessel were performed with toroidally asymmetric disruption loads. Asymmetric halo current conditions were assumed to modify symmetric disruption loads which resulted in net lateral loading on the vacuum vessel torus. Structural analyses with the asymmetric loading indicated significantly higher vessel stress and blanket support forces than with symmetric disruption loads. A recent change in the vessel support design which provided toroidal constraints at each mid port was found to be effective in reducing torus lateral movement and vessel stress

  17. Results from ITER Vacuum Vessel Sector Manufacturing Development in Europe

    International Nuclear Information System (INIS)

    Jones, L.

    2006-01-01

    Significant results have been achieved since the previous SOFT conference, when the manufacturing development work required to prepare for the ITER Vacuum Vessel Sector was described. The contract for the manufacture of a full-size, 20 Ton poloidal part of the inboard section, fabricated according to the ITER reference manufacturing route, including bracing fixtures, welding applications, restraint effects, and fit-up aspects is approaching completion. Since the main aim of the work is to establish the practicability of achieving the tight dimensional tolerances, an accompanying SYSWELD analysis programme has been validation by instrumented welding coupons, and then used for predicting the distortion of the actual construction. A local machining tool has been developed to allow the requirement for machining of the cylindrical features at a late stage of manufacture. Experimental and analytical work has also been carried out to establish the possibility of 3-D cold-forming large sections of walls of the VV. A manufacturing programme to validate an alternative method of fabricating parts of the double-walled VV, utilising e-beam welding only and avoiding the quality issues of the one-sided access and inspection of the closing welds is presented. This paper describes the results of the manufacturing development programme and the future activities. (author)

  18. Manufacturing preparations for the European Vacuum Vessel Sector for ITER

    International Nuclear Information System (INIS)

    Jones, Lawrence; Arbogast, Jean François; Bayon, Angel; Bianchi, Aldo; Caixas, Joan; Facca, Aldo; Fachin, Gianbattista; Fernández, José; Giraud, Benoit; Losasso, Marcello; Löwer, Thorsten; Micó, Gonzalo; Pacheco, Jose Miguel; Paoletti, Roberto; Sanguinetti, Gian Paolo; Stamos, Vassilis; Tacconelli, Massimiliano; Trentea, Alexandru; Utin, Yuri

    2012-01-01

    The contract for the seven European Sectors of the ITER Vacuum Vessel, which has very tight tolerances and high density of welding, was placed at the end of 2010 with AMW, a consortium of three companies. The start-up of the engineering, including R and D, design and analysis activities of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. The statutory and regulatory requirements of ITER Organization and the French Nuclear Safety regulations have made the design development subject to rigorous controls. AMW was able to make use of the previous extensive R and D and prototype work carried out during the past 9 years, especially in relation to advanced welding and inspection techniques. The paper describes the manufacturing methodology with the focus on controlling distortion with predictions by analysis, avoiding use of welded-on jigs, and making use of low heat input narrow-gap welding with electron beam welding as far as possible and narrow-gap TIG when not. Further R and D and more than ten significant mock-ups are described. All these preparations will help to assure the successful manufacture of this critical path item of ITER.

  19. Structural analysis of support structure for ITER vacuum vessel

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka

    2004-12-01

    ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

  20. Design and development of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Koizumi, K.; Nakahira, M.; Itou, Y.; Tada, E. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan); Johnson, G.; Ioki, K.; Elio, F.; Iizuka, T.; Sannazzaro, G.; Takahashi, K.; Utin, Y.; Onozuka, M. [ITER Joint Central Team (JCT), Garching (Germany); Nelson, B. [US Home Team, Oak Ridge National Laboratory (United States); Vallone, C. [EU Home Team, NET Team, Garching (Germany); Kuzmin, E. [RF Home Team, Efremov Institute, City (Russian Federation)

    1998-09-01

    In ITER, the vacuum vessel (VV) is designed to be a water cooled, double-walled toroidal structure made of 316LN stainless steel with a D-shaped cross section approximately 9 m wide and 15 m high. The design work which began at the beginning of the ITER-EDA is nearing completion by resolving the technical issues. In parallel with the design activities, the R and D program, full-scale VV sector model project, was initiated in 1995 to resolve the design and fabrication issues. The full-scale sector model corresponds to an 18 sector (9 sub-sector x 2) and is being fabricated on schedule. To date, 60% of the fabrication had been completed. The fabrication of full-scale model including sector-to-sector connection will be completed by the end of 1997 and performance tests are scheduled until the end of ITER-EDA. This paper describes the latest status of the ITER VV design and the full-scale sector model project. (orig.) 3 refs.

  1. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

    Nakahira, M.; Koizumi, K.; Oka, K.

    2001-01-01

    The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of ± 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

  2. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

    Nakahira, M.; Koizumi, K.; Oka, K.

    1999-01-01

    The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven ITER R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of - 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

  3. Experiments at high elongations in DIII-D

    International Nuclear Information System (INIS)

    Lazarus, E.A.; Turnbull, A.D.; Kellman, A.G.; Ferron, J.R.; Helton, F.J.; Lao, L.L.; Leuer, J.A.; Strait, E.J.; Taylor, T.S.

    1990-01-01

    In this paper we discuss the limitation to elongation observed in D-shaped plasmas in the DIII-D tokamak. We find that as the triangularity is increased and ell i is decreased that the n = O mode takes on an increasingly non-rigid character. Our analysis shows two aspects of the behavior: first, an increasing variation of the m/n = 1/0 component across flux surfaces and second, an increase in the relative amplitude of a m/n = 3/0 component which couples to the m/n = 1/0 component and further destabilizes the mode. In previous work we have reported on study of vertical control and the implementation of those results on DIII-D. In that study we used a single filament, with properties consistent with the radial force balance, to represent the plasma and employed an eigenmode description of the passive shell in order to allow time-ordering of the problem. The most important result of this study was that the active control coil must be positioned in the poloidal plane so as to minimize its interaction with the stabilizing shell currents. As a consequence of plasma toroidicity, these currents flow primarily in the outboard regions of the shell. Thus, control coils on the inboard side of the shell, near the midplane, are required. With such a spatial arrangement we can have radial fields from the active coil penetrating the shell on a time scale faster than the decay of the stabilizing shell currents. In accordance with these model calculations the control system for DIII-D tokamak has been modified resulting in operation to within a few percent of the ideal MHD limit for axisymmetric stability. In this work we refer to the ideal MHD limit as that of the plasma-shell system. The ideal limit can actually be reduced by a poor choice of the active control coils, however that is not the case for work discussed here. 7 refs., 6 figs

  4. Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Lasnier, C. J., E-mail: lasnier@LLNL.gov; Allen, S. L.; Ellis, R. E.; Fenstermacher, M. E.; McLean, A. G.; Meyer, W. H.; Morris, K.; Seppala, L. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551-0808 (United States); Crabtree, K. [College of Optics, University of Arizona, Tucson, Arizona 85721 (United States); Van Zeeland, M. A. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2014-11-15

    An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

  5. Tangles of the ideal separatrix from low mn perturbation in the DIII-D

    Science.gov (United States)

    Goss, Talisa; Crank, Willie; Ali, Halima; Punjabi, Alkesh

    2010-11-01

    The equilibrium EFIT data for the DIII-D shot 115467 at 3000 ms is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates [A. Punjabi, and H. Ali, Phys. Plasmas 15, 122502 (2008); A. Punjabi, Nucl. Fusion 49, 115020 (2009)]. The generating function represents the axisymmetric magnetic geometry and the topology of the DIII-D shot very accurately. A symplectic map for field line trajectories in the natural canonical coordinates in the DIII-D is constructed. We call this map the DIII-D map. The natural canonical coordinates can be readily inverted to physical coordinates (R,φ,Z). Low mn magnetic perturbation with mode numbers (m,n)=(1,1)+(1,-1) is added to the generating function of the map. The amplitude for the low mn perturbation is chosen to be 6X10-4, which is the expected value of the amplitude in tokamaks. The forward and backward DIII-D maps with low mn perturbation are used to calculate the tangles of the ideal separatrix from low mn perturbation in the DIII-D. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  6. Current status of DIII-D real-time digital plasma control

    International Nuclear Information System (INIS)

    Penaflor, B.G.; Piglowski, D.A.; Ferron, J.R.; Walker, M.L.

    1999-06-01

    This paper describes the current status of real-time digital plasma control for the DIII-D tokamak. The digital plasma control system (PCS) has been in place at DIII-D since the early 1990s and continues to expand and improve in its capabilities to monitor and control plasma parameters for DIII-D fusion science experiments. The PCs monitors over 200 tokamak parameters from the DIII-D experiment using a real-time data acquisition system that acquires a new set of samples once every 60 micros. This information is then used in a number of feedback control algorithms to compute and control a variety of parameters including those affecting plasma shape and position. A number of system related improvements has improved the usability and flexibility of the DIII-D PCS. These include more graphical user interfaces to assist in entering and viewing the large and ever growing number of parameters controlled by the PCS, increased interaction and accessibility from other DIII-D applications, and upgrades to the computer hardware and vended software. Future plans for the system include possible upgrades of the real-time computers, further links to other DIII-D diagnostic measurements such as real-time Thomson scattering analysis, and joint collaborations with other tokamak experiments including the NSTX at Princeton

  7. PC application in DIII-D neutral beam operation

    International Nuclear Information System (INIS)

    Gladd, A.S.

    1986-01-01

    An IBM PC/AT has been implemented to improve operation of the DIII-D neutral beams. The PC system provides centralization of all beam data with reasonable access for online shot-to-shot control and analysis. The PC hardware was configured to interface all four neutral beam host mini-computers, support multi-tasking, and provide storage for approximately one month's accumulation of beam data. The PC software is composed of commercial packages used for performance and statistical analysis (i.e. LOTUS 123, PC PLOT, etc.) host communications software (i.e. PCLINK, KERMIT, etc.) and applications developed software utilizing FORTRAN and BASIC. The objectives of this paper are to describe the implementation of the PC system, the methods of integrating the various software packages, and the scenario for online control and analysis

  8. Magnetic Transport Barriers in the DIII-D Tokamak

    Science.gov (United States)

    Kessler, J.; Volpe, F.; Evans, T. E.; Ali, H.; Punjabi, A.

    2009-11-01

    Large overlapping magnetic islands generate chaotic fields. However, a previous work [1] showed that second or third order perturbations of special topology and strength can also generate magnetic diffusion ``barriers" in the middle of stochastic regions. In the present study, we numerically assess their experimental feasibility at DIII-D. For this, realistic I- and C-coils perturbations are superimposed on the equilibrium field and puncture plots are generated with a field-line tracer. A criterion is defined for the automatic recognition of barriers and successfully tested on earlier symplectic maps in magnetic coordinates. The criterion is systematically applied to the new puncture plots in search for dependencies, e.g. upon the edge safety factor q95, which might be relevant to edge localized mode (ELM) stability, as well as to assess the robustness of barriers against fluctuations of the plasma parameters and coil currents. 8pt [1] H. Ali and A. Punjabi, Plasma Phys. Control. Fusion 49, 1565 (2007).

  9. 2-D tomography with bolometry in DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Meyer, W.H.; Geer, B.; Behne, D.M.; Hill, D.N.

    1994-07-01

    We have installed a 48-channel platinum-foil bolometer system on DIII-D achieve better spatial and temporal resolution of the radiated power in diverted discharges. Two 24-channel arrays provide complete plasma coverage with optimized views of the divertor. We have measured the divertor radiation profile for a series of radiative divertor and power balance experiments. We observe a rapid change in the magnitude and distribution of divertor radiation with heavy gas puffing. Unfolding the radiation profile with only two views requires us to treat the core and divertor radiation separately. The core radiation is fitted to a function of magnetic flux and is then subtracted from the divertor viewing chords. The divertor profile is then fit to a 2-D spline as a function of magnetic flux and poloidal angle

  10. Personal computer applications in DIII-D neutral beam operation

    International Nuclear Information System (INIS)

    Glad, A.S.

    1986-01-01

    An IBM PC AT has been implemented to improve operation of the DIII-D neutral beams. The PC system provides centralization of all beam data with reasonable access for on-line shot-to-shot control and analysis. The PC hardware was configured to interface all four neutral beam host minicomputers, support multitasking, and provide storage for approximately one month's accumulation of beam data. The PC software is composed of commercial packages used for performance and statistical analysis (i.e., LOTUS 123, PC PLOT, etc.), host communications software (i.e., PCLink, KERMIT, etc.), and applications developed software utilizing fortran and basIc. The objectives of this paper are to describe the implementation of the PC system, the methods of integrating the various software packages, and the scenario for on-line control and analysis

  11. Physics analysis database for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Schissel, D.P.; Bramson, G.; DeBoo, J.C.

    1986-01-01

    The authors report on a centralized database for handling reduced data for physics analysis implemented for the DIII-D tokamak. Each database record corresponds to a specific snapshot in time for a selected discharge. Features of the database environment include automatic updating, data integrity checks, and data traceability. Reduced data from each diagnostic comprises a dedicated data bank (a subset of the database) with quality assurance provided by a physicist. These data banks will be used to create profile banks which will be input to a transport code to create a transport bank. Access to the database is initially through FORTRAN programs. One user interface, PLOTN, is a command driven program to select and display data subsets. Another user interface, PROF, compares and displays profiles. The database is implemented on a Digital Equipment Corporation VAX 8600 running VMS

  12. Stability in high gain plasmas in DIII-D

    International Nuclear Information System (INIS)

    Lazarus, E.A.; Houlberg, W.A.; Murakami, M.; Wade, M.R.

    1996-10-01

    Fusion power gain has been increased by a factor of 3 in DIII-D plasmas through the use of strong discharge shaping and tailoring of the pressure and current density profiles. H-mode plasmas with weak or negative central magnetic shear are found to have neoclassical ion confinement throughout most of the plasma volume. Improved MHD stability is achieved by controlling the plasma pressure profile width. The highest fusion power gain Q (ratio of fusion power to input power) in deuterium plasmas was 0.0015, which extrapolates to an equivalent Q of 0.32 in a deuterium-tritium plasma and is similar to values achieved in tokamaks of larger size and magnetic fields

  13. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Mahdavi, M.A.; Petrie, T.W.

    1999-02-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. The authors have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. They have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. They have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling they have avoided these and other processes and accessed densities > 1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density

  14. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Baylor, L.R.; Jernigan, T.

    2001-01-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. We have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. We have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. We have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling we have avoided these and other processes and accessed densities >1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density. (author)

  15. Overview of H-mode studies in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Baker, D.R,; Allen, S.L.

    1994-01-01

    A major portion of the DIII-D program includes studies of the L-H transition, of the VH-mode, of particle transport and control and of the power-handling capability of a diverter. Significant progress has been made in all of these areas and the purpose of this paper is to summarize the major results obtained during the last two years. An increased understanding of the origin of improved confinement in H-mode and in VH-mode discharges has been obtained, good impurity control has been achieved in several operating scenarios, studies of helium transport provide encouraging results from the point of view of reactor design, an actively pumped diverter chamber has controlled the density in H-mode discharges and a radiative diverter is a promising technique for controlling the heat flux from the main plasma

  16. Improved edge charge exchange recombination spectroscopy in DIII-D.

    Science.gov (United States)

    Chrystal, C; Burrell, K H; Grierson, B A; Haskey, S R; Groebner, R J; Kaplan, D H; Briesemeister, A

    2016-11-01

    The charge exchange recombination spectroscopy diagnostic on the DIII-D tokamak has been upgraded with the addition of more high radial resolution view chords near the edge of the plasma (r/a > 0.8). The additional views are diagnosed with the same number of spectrometers by placing fiber optics side-by-side at the spectrometer entrance with a precise separation that avoids wavelength shifted crosstalk without the use of bandpass filters. The new views improve measurement of edge impurity parameters in steep gradient, H-mode plasmas with many different shapes. The number of edge view chords with 8 mm radial separation has increased from 16 to 38. New fused silica fibers have improved light throughput and clarify the observation of non-Gaussian spectra that suggest the ion distribution function can be non-Maxwellian in low collisionality plasmas.

  17. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Mahdavi, M.A.; Petrie, T.W.

    1999-01-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. We have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. We have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. We have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling we have avoided these and other processes and accessed densities > 1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density. (author)

  18. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    Leornard, A.W.; Porter, G.D.; Wood, R.D.

    1998-01-01

    The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features

  19. Recent DIII-D neutral beam calibration results

    International Nuclear Information System (INIS)

    Wight, J.; Hong, R.M.; Phillips, J.

    1991-10-01

    Injected DIII-D neutral beam power is estimated based on three principle quantities: the fraction of ion beam that is neutralized in the neutralizer gas cell, the beamline transmission efficiency, and the fraction of beam reionized in the drift duct. System changes in the past few years have included a new gradient grid voltage operating point, ion source arc regulation, routine deuterium operations and new neutralizer gas flow controllers. Additionally, beam diagnostics have been improved and better calibrated. To properly characterize the beams the principle quantities have been re-measured. Two diagnostics are primarily used to measure the quantities. The beamline waterflow calorimetry system measures the neutralization efficiency and the beamline transmission efficiency, and the target tile thermocouples measure the reionization loss. An additional diagnostic, the target tile pyrometer, confirmed the reionization loss measurement. Descriptions and results of these measurements will be presented. 4 refs., 5 figs., 2 tabs

  20. DIII-D ICRF high voltage power supply regulator upgrade

    International Nuclear Information System (INIS)

    Cary, W.P.; Burley, B.L.; Grosnickle, W.H.

    1997-11-01

    For reliable operation and component protection, of the 2 MW 30--120 MHz ICRF Amplifier System on DIII-D, it is desirable for the amplifier to respond to high VSWR conditions as rapidly as possible. This requires a rapid change in power which also means a rapid change in the high voltage power supply current demands. An analysis of the power supply's regulator dynamics was needed to verify its expected operation during such conditions. Based on this information it was found that a new regulator with a larger dynamic range and some anticipation capability would be required. This paper will discuss the system requirements, the as-delivered regulator performance, and the improved performance after installation of the new regulator system. It will also be shown how this improvement has made the amplifier perform at higher power levels more reliably

  1. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

    Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

    1998-07-01

    Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor

  2. The DIII-D Radiative Divertor Project: Status and plans

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Bozek, A.S.

    1996-10-01

    New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots

  3. Engineering design of a Radiative Divertor for DIII-D

    International Nuclear Information System (INIS)

    Smith, J.P.; Allen, S.L.; Anderson, P.M.; Baxi, C.B.; Chin, E.; Fenstermacher, M.E.; Hill, D.N.; Hollerbach, M.A.; Hyatt, A.W.; Junge, R.; Mahdavi, M.A.; Porter, G.D.; Redler, K.; Reis, E.E.; Schaffer, M.J.; Sevier, D.L.; Stambaugh, R.D.

    1995-01-01

    A new divertor called the Radiative Divertor is presently being designed for the DIII-D tokamak. Input from tokamak experiments and modeling form the basis for the new design. The Radiative Divertor is intended to reduce the heat flux on the divertor plates by dispersing the power with radiation. Gas puffing experiments in the current open divertor have shown a reduction of the divertor heat flux with either deuterium or impurity puffing. However, either the plasma density (D 2 ) or the core Z eff (impurities) increases in these experiments. The radiative divertor uses a slot structure to isolate the divertor plasma region from the area surrounding the core plasma. Modeling has shown that the Radiative Divertor hardware will provide better baffling and particle control and thereby minimize the effect of the gas puffing in the divertor region on the plasma core. In addition, the Radiative Divertor structure will allow density control in plasma shapes with high triangularity (>0.8) required for advanced tokamak operation. The divertor structure allows for operation in either double or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. Biasing is an integral part of the design and is based on experience at the Tokamak de Varennes (TdeV) and DIII-D. Boron nitride tiles electrically insulate the inner and outer strike points and a low current electrode is used to apply a radial electric field to the scrape-off layer. TdeV has shown that biasing can provide particle and impurity control. The design is extremely flexible, and will allow physics studies of the effect of slot width and height. This is extremely important, as the amount of chamber volume needed for the divertor in future machines such as International Thermonuclear Experiment Reactor (ITER) and Tokamak Physics Experiment (TPX) must be determined. (orig./WL)

  4. Pedestal performance dependence upon plasma shape in DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Casper, T.A.; Groebner, R.J.; Osborne, T.H.; Snyder, P.B.; Thomas, D.M.

    2007-01-01

    Higher moments of the plasma shape than triangularity are found to significantly affect the pedestal pressure and the edge localized mode (ELM) characteristics in DIII-D. The shape dependence of the pedestal pressure was experimentally examined by varying the squareness in the proposed ITER configuration while holding the triangularity fixed. Over this scan the pedestal pressure increased by ∼50% from highest squareness to lowest squareness. The variation of pedestal energy is found to be consistent with the stability analysis of the measured profiles. The ELM energy also varied with the shape to maintain a nearly constant fraction of the pedestal energy. Stability analysis using model shapes and pressure profiles indicates that much of the advantage of high triangularity for high pedestal pressure can be achieved in lower triangularity shapes by optimizing squareness and/or the distance of the secondary upper separatrix from the primary separatrix. In high beta discharges an increase in pedestal pressure is observed with higher global stored energy. The greatest pedestal pressure increase is at low squareness due to an increase in both the pressure gradient stability limit and the width of the pedestal. The variation in pedestal pressure with squareness was also used to optimize 'hybrid' discharges in DIII-D where a lower pedestal pressure was required for an improved overall performance. In the 'hybrid' regime low squareness resulted in a high pedestal pressure with large infrequent ELMs that eventually triggered an internal 2/1 tearing mode that locked, resulting in a disruption. At higher squareness the pedestal pressure was reduced with smaller and more rapid ELMs, resulting in the maintenance of a steady beneficial internal 3/2 tearing mode and good confinement. For all the cases studied, an increase in the pedestal width at low squareness appears to be a significant factor in the increase in the total pedestal pressure

  5. Analysis of toroidal vacuum vessels for use in demonstration sized tokamak reactors

    International Nuclear Information System (INIS)

    Culbert, M.E.

    1978-07-01

    The vacuum vessel component of the tokamak fusion reactor is the subject of this study. The main objective of this paper was to provide guidance for the structural design of a thin wall externally pressurized toroidal vacuum vessel. The analyses are based on the available state-of-the-art analytical methods. The shortcomings of these analytical methods necessitated approximations and assumptions to be made throughout the study. A principal result of the study has been the identification of a viable vacuum vessel design for the Demonstration Tokamak Hybrid Reactor (DTHR) and The Next Step (TNS) Reactor

  6. The upgrade of the DIII-D EC system using 120 GHz ITER gyrotrons

    International Nuclear Information System (INIS)

    Callis, R.W.; Lohr, J.; Gorelov, I.A.; Ponce, D.; Kajiwara, K.; Tooker, J.F.

    2005-01-01

    The planned growth in the EC system on DIII-D over the next few years requires the installation of two depressed collector gyrotrons, a high voltage power supply, two low loss transmission lines, and the required support equipment. This new DIII-D EC equipment could be made identical to the ITER EC system requirements. By building the DIII-D hardware to the ITER specifications, it will allow ITER to gain beneficial prototyping experience on a working tokamak, prior to committing to building the hardware for delivery to ITER

  7. A system to deposit boron films (boronization) in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Hodapp, T.R.; Jackson, G.L.; Phillips, J.; Holtrop, K.L.; Peterson, P.L.; Winters, J.

    1992-01-01

    A system has been added to the DIII-D tokamak to coat its plasma facing surfaces with a film of boron using diborane gas. The system includes special health and safety equipment for handling the diborane gas which is toxic and inflammable. The purpose f the boron film is to reduce the levels of impurity atoms in the DIII-D plasmas. Experiments following the application of the boron film in DIII-D have led to significant reductions in plasma impurity levels and the observation of a new, very high confinement regime

  8. The importance of matched poloidal spectra to error field correction in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Paz-Soldan, C., E-mail: paz-soldan@fusion.gat.com; Lanctot, M. J.; Buttery, R. J.; La Haye, R. J.; Strait, E. J. [General Atomics, P.O. Box 85608, San Diego, California 92121 (United States); Logan, N. C.; Park, J.-K.; Solomon, W. M. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Shiraki, D.; Hanson, J. M. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2014-07-15

    Optimal error field correction (EFC) is thought to be achieved when coupling to the least-stable “dominant” mode of the plasma is nulled at each toroidal mode number (n). The limit of this picture is tested in the DIII-D tokamak by applying superpositions of in- and ex-vessel coil set n = 1 fields calculated to be fully orthogonal to the n = 1 dominant mode. In co-rotating H-mode and low-density Ohmic scenarios, the plasma is found to be, respectively, 7× and 20× less sensitive to the orthogonal field as compared to the in-vessel coil set field. For the scenarios investigated, any geometry of EFC coil can thus recover a strong majority of the detrimental effect introduced by the n = 1 error field. Despite low sensitivity to the orthogonal field, its optimization in H-mode is shown to be consistent with minimizing the neoclassical toroidal viscosity torque and not the higher-order n = 1 mode coupling.

  9. Real-time, vibration-compensated CO2 interferometer operation on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Ahlgren, D.R.; Crosbie, J.

    1988-01-01

    A multichannel, two-color, quadrature heterodyne interferometer is used to measure the line density in the DIII-D tokamak. The unique feature of this real-time vibration-compensated interferometer is the combination of high speed (1 MHz), high resolution (2π/256), and wide range ( +- 8193 fringes). Quadrature phase information from a CO 2 laser (10.6 μm) and a He--Ne laser (0.63 μm) are digitized with high-speed (6 MHz) flash digitizers. Zero crossings of the signals are counted with digital circuitry yielding quarter fringe resolution with a 4-MHz bandwidth. Further fringe resolution of 1/256 is provided at 350 kHz by a PROM which uses the digital signals as input to a look-up table. Analog line density is presently available at 80 kHz with a system noise equivalent phase shift of +- 2/256. Error monitoring is provided for low signal amplitude and exceeding the maximum fringe rate. In addition, a method to prevent coating of in-vessel mirrors due to plasma and vessel wall cleaning discharges has been developed

  10. Recovery process of wall condition in KSTAR vacuum vessel after temporal machine-vent for repair

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwang Pyo, E-mail: kpkim@nfri.er.ke; Hong, Suk-Ho; Lee, Hyunmyung; Song, Jae-in; Jung, Nam-Yong; Lee, Kunsu; Chu, Yong; Kim, Hakkun; Park, Kaprai; Oh, Yeong-Kook

    2015-10-15

    Highlights: • Efforts have been made to obtain vacuum condition that is essential for the plasma experiments. • For example, the vacuum vessel should be vented to repair in-vessel components such as diagnostic shutter, and PFC damaged by high energy plasma. • Here, we present the recovery process of wall condition in KSTAR after temporal machine-vent for repair. • It is found that an acceptable vacuum condition has been achieved only by plasma based wall conditioning techniques such as baking, GDC, and boronization. • This study was that the proper recovering method of the vacuum condition should be developed according to the severity of the accident. - Abstract: Efforts have been made to obtain vacuum condition that is essential for the plasma experiments. Under certain situations, for example, the vacuum vessel should be vented to repair in-vessel components such as diagnostic shutter, exchange of window for diagnostic equipment, and PFC damaged by high energy plasma. For the quick restart of the campaign, a recovery process was established to make the vacuum condition acceptable for the plasma experiment. In this paper, we present the recovery process of wall condition in KSTAR after temporal machine-vent for repair. It is found that an acceptable vacuum condition has been achieved only by plasma based wall conditioning techniques such as baking, GDC, and boronization. This study was that the proper recovering method of the vacuum condition should be developed according to the severity of the accident.

  11. Design description of the vacuum vessel for the Advanced Toroidal Facility

    International Nuclear Information System (INIS)

    Chipley, K.K.; Nelson, B.E.; Vinyard, L.M.; Williamson, D.F.

    1983-01-01

    The Advanced Toroidal Facility (ATF) will be a stellarator experiment to investigate improvements in toroidal confinement. The vacuum vessel for this facility will provide the appropriate evacuated region for plasma containment within the helical field (HF) coils. The vessel is designed to provide the maximum reasonable volume inside the HF coils and to provide the maximum reasonable access for future diagnostics. The vacuum vessel design is at an early phase and all of the details have not been completed. The heat transfer analysis and stress analysis completed during the conceptual design indicate that the vessel will not change drastically

  12. Model-based dynamic resistive wall mode identification and feedback control in the DIII-D tokamak

    International Nuclear Information System (INIS)

    In, Y.; Kim, J.S.; Edgell, D.H.; Strait, E.J.; Humphreys, D.A.; Walker, M.L.; Jackson, G.L.; Chu, M.S.; Johnson, R.; La Haye, R.J.; Okabayashi, M.; Garofalo, A.M.; Reimerdes, H.

    2006-01-01

    A new model-based dynamic resistive wall mode (RWM) identification and feedback control algorithm has been developed. While the overall RWM structure can be detected by a model-based matched filter in a similar manner to a conventional sensor-based scheme, it is significantly influenced by edge-localized-modes (ELMs). A recent study suggested that such ELM noise might cause the RWM control system to respond in an undesirable way. Thus, an advanced algorithm to discriminate ELMs from RWM has been incorporated into this model-based control scheme, dynamic Kalman filter. Specifically, the DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] resistive vessel wall was modeled in two ways: picture frame model or eigenmode treatment. Based on the picture frame model, the first real-time, closed-loop test results of the Kalman filter algorithms during DIII-D experimental operation are presented. The Kalman filtering scheme was experimentally confirmed to be effective in discriminating ELMs from RWM. As a result, the actuator coils (I-coils) were rarely excited during ELMs, while retaining the sensitivity to RWM. However, finding an optimized set of operating parameters for the control algorithm requires further analysis and design. Meanwhile, a more advanced Kalman filter based on a more accurate eigenmode model has been developed. According to this eigenmode approach, significant improvement in terms of control performance has been predicted, while maintaining good ELM discrimination

  13. Recent DIII-D high power heating and current drive experiments

    International Nuclear Information System (INIS)

    Simonen, T.C.; Jackson, G.L.; Lazarus, E.A.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.

    1995-01-01

    This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal β. ((orig.))

  14. DIII-D research operations. Annual report, October 1, 1991--September 30, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D. [ed.

    1993-05-01

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R&D; and collaborative efforts.

  15. Recent DIII-D high power heating and current drive experiments

    International Nuclear Information System (INIS)

    Simonen, T.C.; Jackson, G.L.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.; Lazarus, E.A.

    1994-02-01

    This paper describes recent DIII-D high power heating and current drive experiments. Describes are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal beta

  16. Recent DIII-D high power heating and current drive experiments

    Energy Technology Data Exchange (ETDEWEB)

    Simonen, T.C. [General Atomics, San Diego, CA (United States); Jackson, G.L. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Politzer, P.A. [General Atomics, San Diego, CA (United States); Taylor, T.S. [General Atomics, San Diego, CA (United States); DIII-D Team

    1995-01-01

    This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal {beta}. ((orig.)).

  17. Engineering, installation, testing, and initial operation of the DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

    Andersen, P.M.; Baxi, C.B.; Reis, E.E.; Schaffer, M.J.; Smith, J.P.

    1990-09-01

    The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint. 2 refs., 5 figs

  18. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  19. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

  20. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  1. Real time software for the control and monitoring of DIII-D system interlocks

    International Nuclear Information System (INIS)

    Broesch, J.D.; Penaflor, B.G.; Coon, R.M.; Harris, J.J.; Scoville, J.T.

    1996-10-01

    This paper describes the real time, multi-tasking, multi-user software and communications of the E-Power Supply System Integrated Controller (EPSSIC) for the DIII-D tokamak. EPSSIC performs the DIII-D system wide go/no-go determination for the plasma sequencing. This paper discusses the data module handling, task work load balancing, and communications requirements. Operational experience with the new EPSSIC and recent improvements to this system are also described

  2. Enhanced Computational Infrastructure for Data Analysis at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schissel, D.P.; Peng, Q.; Schachter, J.; Tepstra, T.B.; Casper, T.A.; Freeman, J.; Jong, R.; Keith, K.M.; McHarg, B.B. Jr; Meyer, W.H.; Parker, C.T.; Warner, A.M.

    1999-01-01

    The DIII-D National Team consists of about 120 operating staff and 100 research scientists drawn from 9 U.S. National Laboratories, 19 foreign laboratories, 16 universities, and 5 industrial partnerships. This multi-institution collaboration carries out the integrated DIII-D program mission which is to establish the scientific basis for the optimization of the tokamak approach to fusion energy production. Presently, about two-thirds of the research physics staff are from the national and international collaborating institutions

  3. Enhanced computational infrastructure for data analysis at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schissel, D.P.; Peng, Q.; Schachter, J.; Terpstra, T.B.; Casper, T.A.; Freeman, J.; Jong, R.; Keith, K.M.; McHarg, B.B.; Meyer, W.H.; Parker, C.T.

    2000-01-01

    Recently a number of enhancements to the computer hardware infrastructure have been implemented at the DIII-D National Fusion Facility. Utilizing these improvements to the hardware infrastructure, software enhancements are focusing on streamlined analysis, automation, and graphical user interface (GUI) systems to enlarge the user base. The adoption of the load balancing software package LSF Suite by Platform Computing has dramatically increased the availability of CPU cycles and the efficiency of their use. Streamlined analysis has been aided by the adoption of the MDSplus system to provide a unified interface to analyzed DIII-D data. The majority of MDSplus data is made available in between pulses giving the researcher critical information before setting up the next pulse. Work on data viewing and analysis tools focuses on efficient GUI design with object-oriented programming (OOP) for maximum code flexibility. Work to enhance the computational infrastructure at DIII-D has included a significant effort to aid the remote collaborator since the DIII-D National Team consists of scientists from nine national laboratories, 19 foreign laboratories, 16 universities, and five industrial partnerships. As a result of this work, DIII-D data is available on a 24x7 basis from a set of viewing and analysis tools that can be run on either the collaborators' or DIII-D's computer systems. Additionally, a web based data and code documentation system has been created to aid the novice and expert user alike

  4. Effect of heating scheme on SOL width in DIII-D and EAST

    Directory of Open Access Journals (Sweden)

    L. Wang

    2017-08-01

    Full Text Available Joint DIII-D/EAST experiments in the radio-frequency (RF heated H-mode scheme with comparison to that of neutral beam (NB heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH H-mode plasmas.

  5. Radiation asymmetries during disruptions on DIII-D caused by massive gas injection

    International Nuclear Information System (INIS)

    Commaux, N.; Baylor, L. R.; Jernigan, T. C.; Foust, C. R.; Combs, S.; Meitner, S. J.; Hollmann, E. M.; Izzo, V. A.; Moyer, R. A.; Humphreys, D. A.; Wesley, J. C.; Eidietis, N. W.; Parks, P. B.; Lasnier, C. J.

    2014-01-01

    One of the major challenges that the ITER tokamak will have to face during its operations are disruptions. During the last few years, it has been proven that the global consequences of a disruption can be mitigated by the injection of large quantities of impurities. But one aspect that has been difficult to study was the possibility of local effects inside the torus during such injection that could damage a portion of the device despite the global heat losses and generated currents remaining below design parameter. 3D MHD simulations show that there is a potential for large toroidal asymmetries of the radiated power during impurity injection due to the interaction between the particle injection plume and a large n = 1 mode. Another aspect of 3D effects is the potential occurrence of Vertical Displacement Events (VDE), which could induce large poloidal heat load asymmetries. This potential deleterious effect of 3D phenomena has been studied on the DIII-D tokamak, thanks to the implementation of a multi-location massive gas injection (MGI) system as well as new diagnostic capabilities. This study showed the existence of a correlation between the location of the n = 1 mode and the local heat load on the plasma facing components but shows also that this effect is much smaller than anticipated (peaking factor of ∼1.1 vs 3-4 according to the simulations). There seems to be no observable heat load on the first wall of DIII-D at the location of the impurity injection port as well as no significant radiation asymmetries whether one or 2 valves are fired. This study enabled the first attempt of mitigation of a VDE using impurity injection at different poloidal locations. The results showed a more favorable heat deposition when the VDE is mitigated early (right at the onset) by impurity injection. No significant improvement of the heat load mitigation efficiency has been observed for late particle injection whether the injection is done “in the way” of the VDE

  6. Radiation asymmetries during disruptions on DIII-D caused by massive gas injectiona)

    Science.gov (United States)

    Commaux, N.; Baylor, L. R.; Jernigan, T. C.; Hollmann, E. M.; Humphreys, D. A.; Wesley, J. C.; Izzo, V. A.; Eidietis, N. W.; Lasnier, C. J.; Moyer, R. A.; Parks, P. B.; Foust, C. R.; Combs, S.; Meitner, S. J.

    2014-10-01

    One of the major challenges that the ITER tokamak will have to face during its operations are disruptions. During the last few years, it has been proven that the global consequences of a disruption can be mitigated by the injection of large quantities of impurities. But one aspect that has been difficult to study was the possibility of local effects inside the torus during such injection that could damage a portion of the device despite the global heat losses and generated currents remaining below design parameter. 3D MHD simulations show that there is a potential for large toroidal asymmetries of the radiated power during impurity injection due to the interaction between the particle injection plume and a large n = 1 mode. Another aspect of 3D effects is the potential occurrence of Vertical Displacement Events (VDE), which could induce large poloidal heat load asymmetries. This potential deleterious effect of 3D phenomena has been studied on the DIII-D tokamak, thanks to the implementation of a multi-location massive gas injection (MGI) system as well as new diagnostic capabilities. This study showed the existence of a correlation between the location of the n = 1 mode and the local heat load on the plasma facing components but shows also that this effect is much smaller than anticipated (peaking factor of ˜1.1 vs 3-4 according to the simulations). There seems to be no observable heat load on the first wall of DIII-D at the location of the impurity injection port as well as no significant radiation asymmetries whether one or 2 valves are fired. This study enabled the first attempt of mitigation of a VDE using impurity injection at different poloidal locations. The results showed a more favorable heat deposition when the VDE is mitigated early (right at the onset) by impurity injection. No significant improvement of the heat load mitigation efficiency has been observed for late particle injection whether the injection is done "in the way" of the VDE (upward VDE

  7. A new gamma ray imaging diagnostic for runaway electron studies at DIII-D

    Science.gov (United States)

    Cooper, C. M.; Pace, D. C.; Eidietis, N. W.; Paz-Soldan, C.; Commaux, N.; Shiraki, D.; Hollmann, E. M.; Moyer, R. A.; Risov, V.

    2015-11-01

    A new Gamma Ray Imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at DIII-D. The diagnostic is sensitive to 0.5 - 50 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE dissipation from pellet injection. The GRI consists of a lead ``pinhole camera'' mounted on the midplane with 11x11 counter-current tangential chords 20 cm wide that span the vessel. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE Bremsstrahlung radiation. Detectors operate in current saturation mode at 10 MHz, or the flux is attenuated for Pulse Height Analysis (PHA) capable of discriminating up to ~10k pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. Work supported by the US DOE under DE-AC05-00OR22725, DE-FG02-07ER54917 & DE-FC02-04ER54698.

  8. Optimization of DIII-D discharges to avoid AE destabilization

    Science.gov (United States)

    Varela, Jacobo; Spong, Donald; Garcia, Luis; Huang, Juan; Murakami, Masanori

    2017-10-01

    The aim of the study is to analyze the stability of Alfven Eigenmodes (AE) perturbed by energetic particles (EP) during DIII-D operation. We identify the optimal NBI operational regimes that avoid or minimize the negative effects of AE on the device performance. We use the reduced MHD equations to describe the linear evolution of the poloidal flux and the toroidal component of the vorticity in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles, including the effect of the acoustic modes. We add the Landau damping and resonant destabilization effects using a closure relation. We perform parametric studies of the MHD and AE stability, taking into account the experimental profiles of the thermal plasma and EP, also using a range of values of the energetic particles β, density and velocity as well the effect of the toroidal couplings. We reproduce the AE activity observed in high poloidal β discharge at the pedestal and reverse shear discharges. This material based on work is supported both by the U.S. Department of Energy, Office of Science, under Contract DE-AC05-00OR22725 with UT-Battelle, LLC. Research sponsored in part by the Ministerio de Economia y Competitividad of Spain under the project.

  9. The DIII-D Tokamak trouble report database

    International Nuclear Information System (INIS)

    Petersen, P.I.; Miller, S.M.

    1992-01-01

    Operation of the DIII-D tokamak at General Atomics 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 is used to find the major problem areas so they can be solved and improve the tokamak availabilty. The data is available to Idaho National Engineering Laboratory (INEL). They are 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. In this paper, the authors will discuss these failures and the actions taken to correct them. The data base also will be used to determine the preventive maintenance schedule for different components

  10. The DIII-D 3 MW, 110 GHz ECH System

    International Nuclear Information System (INIS)

    Callis, R.W.; Lohr, J.; Ponce, D.; O'Neill, R.C.; Prater, R.; Luce, T.C.

    1999-01-01

    Three 110 GHz gyrotrons with nominal output power of 1 MW each have been installed and are operational on the DIII-D tokamak. One gyrotron is built by Gycom and has a nominal rating of 1 MW and a 2 s pulse length, with the pulse length being determined by the maximum temperature allowed on the edge cooled Boron Nitride window. The second and third gyrotrons were built by Communications and Power Industries (CPI). The first CPI gyrotron uses a double disc FC-75 cooled sapphire window which has a pulse length rating of 0.8 s at 1 MW, 2s at 0.5 MW and 10s at 0.35 MW. The second CPI gyrotron, utilizes a single disc chemical-vapor-deposition diamond window, that employs water cooling around the edge of the disc. Calculation predict that the diamond window should be capable of full 1 MW cw operation. All gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HEl 1 mode. Each waveguide system incorporates a two mirror launcher which can steer the rf beam poloidally from the center to the outer edge of the plasma. Central current drive experiments with the two gyrotrons with 1.5 MW of injected power drove about 0.17 MA. Results from using the three gyrotron systems will be reported as well as the plans to upgrade the system to 6 MW

  11. Computerized operation of the DIII-D neutral beams

    International Nuclear Information System (INIS)

    Glad, A.S.; Tooker, J.F.

    1986-01-01

    Operation of the DIII-D neutral beams utilizes computerized control to provide routine tokamak beam heating shots and an effective method for automatic ion source operation. Computerized control reduces operational complexity, thus providing consistent reliability and availability of beams and a significant reduction in the the costs of routine operation. The objectives in implementing computerized control for operation were: (1) to improve operator efficiency for controlling multiple beam lines and increasing beam availability through standard procedures, (2) to provide a simplified scheme that operators and coordinators can construct and maintain, and (3) to provide a single integrated mechanism for both tokamak operation and automatic source conditioning. The years of experience in operating neutral beams at Doublet III provided the data necessary to meet the objectives. The method for computerized control consisted of three integrated functions: (1) a structured command language was implemented to provide the mechanism for automatically sequencing beams, (2) a historical file was constructed from the operational parameters to characterize the ion source, and consists of data from approximately 100,000 beam shots, and (3) procedures were developed integrating the language to the historical file for normal operation and source conditioning. This paper describes the method for sequencing beams automatically, the structure of the historical data file, and the procedures which integrate the historical data with tokamak operation and automatic source conditioning

  12. Fast wave current drive in DIII-D

    International Nuclear Information System (INIS)

    Petty, C.C.; Callis, R.W.; Chiu, S.C.; deGrassie, J.S.; Forest, C.B.; Freeman, R.L.; Gohil, P.; Harvey, R.W.; Ikezi, H.; Lin-Liu, Y.-R.

    1995-02-01

    The non-inductive current drive from fast Alfven waves launched by a directional four-element antenna was measured in the DIII-D tokamak. The fast wave frequency (60 MHz) was eight times the deuterium cyclotron frequency at the plasma center. An array of rf pickup loops at several locations around the torus was used to verify the directivity of the four-element antenna. Complete non-inductive current drive was achieved using a combination of fast wave current drive (FWCD) and electron cyclotron current drive (ECCD) in discharges for which the total plasma current was inductively ramped down from 400 to 170 kA. For discharges with steady plasma current, up to 110 kA of FWCD was inferred from an analysis of the loop voltage, with a maximum non-inductive current (FWCD, ECCD, and bootstrap) of 195 out of 310 kA. The FWCD efficiency increased linearly with central electron temperature. For low current discharges, the FWCD efficiency was degraded due to incomplete fast wave damping. The experimental FWCD was found to agree with predictions from the CURRAY ray-tracing code only when a parasitic loss of 4% per pass was included in the modeling along with multiple pass damping

  13. Electron cyclotron current drive experiments on DIII-D

    International Nuclear Information System (INIS)

    James, R.A.; Giruzzi, G.; Gentile, B. de; Rodriguez, L.; Fyaretdinov, A.; Gorelov, Yu.; Trukhin, V.; Harvey, R.; Lohr, J.; Luce, T.C.; Matsuda, K.; Politzer, P.; Prater, R.; Snider, R.; Janz, S.

    1990-05-01

    Electron Cyclotron Current Drive (ECCD) experiments on the DIII-D tokamak have been performed using 60 GHz waves launched from the high field side of the torus. Preliminary analysis indicates rf driven currents between 50 and 100 kA in discharges with total plasma currents between 200 and 500 kA. These are the first ECCD experiments with strong first pass absorption, localized deposition of the rf power, and τ E much longer than the slowing-down time of the rf generated current carriers. The experimentally measured profiles for T e , η e and Z eff are used as input for a 1D transport code and a multiply-ray, 3D ray tracing code. Comparisons with theory and assessment of the influence of the residual electric field, using a Fokker-Planck code, are in progress. The ECH power levels were between 1 and 1.5 MW with pulse lengths of about 500 msec. ECCD experiments worldwide are motivated by issues relating to the physics and technical advantages of the use of high frequency rf waves to drive localized currents. ECCD is accomplished by preferentially heating electrons moving in one toroidal direction, reducing their collisionality and thereby producing a non-inductively driven toroidal current. 6 refs., 4 figs

  14. Simulation of DIII-D Flat q Discharges

    International Nuclear Information System (INIS)

    Kessel, C.E.; Garofalo, A.; Terpstra, T.

    2004-01-01

    The Advanced Tokamak plasma configuration has significant potential for the economical production of fusion power. Research on various tokamak experiments are pursuing these plasmas to establish high β, high bootstrap current fraction, 100% noninductive current, and good energy confinement, in a quasi-stationary state. One candidate is the flat q discharge produced in DIII-D, where the safety factor varies from 2.0 on axis, to slightly below 2.0 at the minimum, and then rises to about 3.5 at the 95% surface. This plasma is prototypical of those studied for power plants in the ARIES tokamak studies. The plasma is produced by ramping up the plasma current and ramping down the toroidal field throughout the discharge. The plasma current reaches 1.65 MA, and the toroidal field goes from 2.25 to 1.6 T. The q min remains high and at large radius, ρ ∼ 0.6. The plasma establishes an internal transport barrier in the ion channel, and transitions to H-mode. The free-boundary Tokamak Simulation Code (TSC) is being used to model the discharge and project the impact of changes in the plasma current, toroidal field, and injected power programming

  15. Gamma ray imager on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.; Van Zeeland, M. A.; Watkins, M. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Cooper, C. M. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Riso, V. [State University of New York-Buffalo, 12 Capen Hall, Buffalo, New York 14260-1660 (United States)

    2016-04-15

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1–60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

  16. HIGH PERFORMANCE STATIONARY DISCHARGES IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    Luce, T.C.; Wade, M.R.; Ferron, J.R.; Politzer, P.A.; Hyatt, A.W.; Sips, A.C.C.; Murakami, M.

    2003-01-01

    Recent experiments in the DIII-D tokamak [J.L. Luxon, Nucl. Fusion 42,614 (2002)] have demonstrated high β with good confinement quality under stationary conditions. Two classes of stationary discharges are observed--low q 95 discharges with sawteeth and higher q 95 without sawteeth. The discharges are deemed stationary when the plasma conditions are maintained for times greater than the current profile relaxation time. In both cases the normalized fusion performance (β N H 89P /q 95 2 ) reaches or exceeds the value of this parameter projected for Q fus = 10 in the International Thermonuclear Experimental Reactor (ITER) design [R. Aymar, et al., Plasma Phys. Control. Fusion 44, 519 (2002)]. The presence of sawteeth reduces the maximum achievable normalized β, while confinement quality (confinement time relative to scalings) is largely independent of q 95 . Even with the reduced β limit, the normalized fusion performance maximizes at the lowest q 95 . Projections to burning plasma conditions are discussed, including the methodology of the projection and the key physics issues which still require investigation

  17. Overview of Recent DIII-D Experimental Results

    Science.gov (United States)

    Fenstermacher, Max

    2015-11-01

    Recent DIII-D experiments have added to the ITER physics basis and to physics understanding for extrapolation to future devices. ELMs were suppressed by RMPs in He plasmas consistent with ITER non-nuclear phase conditions, and in steady state hybrid plasmas. Characteristics of the EHO during both standard high torque, and low torque enhanced pedestal QH-mode with edge broadband fluctuations were measured, including edge localized density fluctuations with a microwave imaging reflectometer. The path to Super H-mode was verified at high beta with a QH-mode edge, and in plasmas with ELMs triggered by Li granules. ITER acceptable TQ mitigation was obtained with low Ne fraction Shattered Pellet Injection. Divertor ne and Te data from Thomson Scattering confirm predicted drift-driven asymmetries in electron pressure, and X-divertor heat flux reduction and detachment were characterized. The crucial mechanisms for ExB shear control of turbulence were clarified. In collaboration with EAST, high beta-p scenarios were obtained with 80 % bootstrap fraction, high H-factor and stability limits, and large radius ITBs leading to low AE activity. Work supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  18. ELM-Induced Plasma Wall Interactions in DIII-D

    International Nuclear Information System (INIS)

    Rudakov, D.L.; Boedo, J.A.; Yu, J.H.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Hollmann, E.M.; Lasnier, C.J.; McLean, A.G.; Moyer, R.A.; Stangeby, P.C.; Tynan, G.R.; Wampler, W.R.; Watkins, J.G.; West, W.P.; Wong, C.C.; Zeng, L.; Bastasz, R.J.; Buchenauer, D.; Whaley, J.

    2008-01-01

    Intense transient fluxes of particles and heat to the main chamber components induced by edge localized modes (ELMs) are of serious concern for ITER. In DIII-D, plasma interaction with the outboard chamber wall is studied using Langmuir probes and optical diagnostics including a fast framing camera. Camera data shows that ELMs feature helical filamentary structures localized at the low field side of the plasma and aligned with the local magnetic field. During the nonlinear phase of an ELM, multiple filaments are ejected from the plasma edge and propagate towards the outboard wall with velocities of 0.5-0.7 km/s. When reaching the wall, filaments result in 'hot spots'--regions of local intense plasma-material interaction (PMI) where the peak incident particle and heat fluxes are up to 2 orders of magnitude higher than those between ELMs. This interaction pattern has a complicated geometry and is neither toroidally nor poloidally symmetric. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. In high density/collisionality discharges, contributions of ELMs and inter-ELM periods to PMI at the wall are comparable. A Midplane Material Evaluation Station (MiMES) has been recently installed in order to conduct in situ measurements of erosion/redeposition at the outboard chamber wall, including those caused by ELMs

  19. An overview of the DIII-D program

    International Nuclear Information System (INIS)

    Luxon, J.L.

    1996-10-01

    The DIII-D program focuses on developing fusion physics in an integrated program of tokamak concept improvement. The intent is both to support the present ITER physics R and D and to develop more efficient concepts for the later phases of ITER and eventual power plants. Progress in this effort can be best summarized by recent results for a diverted deuterium discharge with negative central shear which reached a performance level of Q DT = 0.32. The ongoing development of the tools needed to carry out this program of understanding and optimization continues to be crucial to its success. Control of the plasma cross-sectional shape and the internal distributions of plasma current, density, and rotation has been essential to optimizing plasma performance. Advanced divertor concepts provide edge power and particle control for future devices such as ITER and provide techniques to help manage the edge power and particle flows for advanced tokamak concepts. New divertor diagnostics and improved modeling are developing excellent divertor understanding. Many of the plasma physics issues being posed by ITER are being addressed. Scrapeoff layer power flow is being characterized to provide an accurate basis for the design of reactor devices. Ongoing studies of the density limit focus on identifying ways in which ITER can achieve the required densities in excess of the Greenwald limit. Better understanding of disruptions is crucial to the design of future reactors

  20. Electron cyclotron current drive experiments on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    James, R.A. (Lawrence Livermore National Lab., CA (USA)); Giruzzi, G.; Gentile, B. de; Rodriguez, L. (Association Euratom-CEA, Centre d' Etudes Nucleaires de Cadarache, 13 - Saint-Paul-les-Durance (France)); Fyaretdinov, A.; Gorelov, Yu.; Trukhin, V. (Kurchatov Inst. of Atomic Energy, Moscow (USSR)); Harvey, R.; Lohr, J.; Luce, T.C.; Matsuda, K.; Politzer, P.; Prater, R.; Snider, R. (General Atomics, San Di

    1990-05-01

    Electron Cyclotron Current Drive (ECCD) experiments on the DIII-D tokamak have been performed using 60 GHz waves launched from the high field side of the torus. Preliminary analysis indicates rf driven currents between 50 and 100 kA in discharges with total plasma currents between 200 and 500 kA. These are the first ECCD experiments with strong first pass absorption, localized deposition of the rf power, and {tau}{sub E} much longer than the slowing-down time of the rf generated current carriers. The experimentally measured profiles for T{sub e}, {eta}{sub e} and Z{sub eff} are used as input for a 1D transport code and a multiply-ray, 3D ray tracing code. Comparisons with theory and assessment of the influence of the residual electric field, using a Fokker-Planck code, are in progress. The ECH power levels were between 1 and 1.5 MW with pulse lengths of about 500 msec. ECCD experiments worldwide are motivated by issues relating to the physics and technical advantages of the use of high frequency rf waves to drive localized currents. ECCD is accomplished by preferentially heating electrons moving in one toroidal direction, reducing their collisionality and thereby producing a non-inductively driven toroidal current. 6 refs., 4 figs.

  1. State transitions, hysteresis, and control parameters on DIII-D

    International Nuclear Information System (INIS)

    Thomas, D.M.; Groebner, R.J.; Carlstrom, T.N.; Osborne, T.H.; Petrie, T.W.

    1998-07-01

    The theory of turbulence decorrelation by ExB velocity shear is the leading candidate to explain the changes in turbulence and transport that are seen at the plasma edge at the L to H transition. Based on this, a key question is: What are the conditions or control parameters needed to begin the formation of the E r shear layer and thus trigger the L to H transition? On the DIII-D tokamak, the authors are attacking this question both through direct tests of the various theories and by trying to gain insight into the fundamental physics by investigating the control parameters which have a major effect on the power threshold. In this paper the authors describe results of studies on oscillating discharges where the plasma transitions continuously between L and H states. By following the dynamics of the plasma state through the forward and back transitions, they can represent the evolution of various control parameter candidates as a trajectory in various parametric spaces. The shape of these control curves can illustrate the specific nonlinearities governing the L-H transition problem, and under the proper conditions may be interpreted in the context of various phase-transition based models. In particular, the hysteresis exhibited in the various curves may help to clarify causality (what are the critical parameters) and may serve as tests of the models, given sufficient experimental accuracy. At present they are looking at T e , E r and ballooning/diamagnetic parameters as possible control parameter candidates

  2. ELMs IN DIII-D HIGH PERFORMANCE DISCHARGES

    International Nuclear Information System (INIS)

    TURNBULL, A.D; LAO, L.L; OSBORNE, T.H; SAUTER, O; STRAIT, E.J; TAYLOR, T.S; CHU, M.S; FERRON, J.R; GREENFIELD, C.M; LEONARD, A.W; MILLER, R.L; SNYDER, P.B; WILSON, H.R; ZOHM, H

    2003-01-01

    A new understanding of edge localized modes (ELMs) in tokamak discharges is emerging [P.B. Snyder, et al., Phys. Plasmas, 9, 2037 (2002)], in which the ELM is an essentially ideal magnetohydrodynamic (MHD) instability and the ELM severity is determined by the radial width of the linearly unstable MHD kink modes. A detailed, comparative study of the penetration into the core of the respective linear instabilities in a standard DIII-D ELMing, high confinement mode (H-mode) discharge, with that for two relatively high performance discharges shows that these are also encompassed within the framework of the new model. These instabilities represent the key, limiting factor in extending the high performance of these discharges. In the standard ELMing H-mode, the MHD instabilities are highly localized in the outer few percent flux surfaces and the ELM is benign, causing only a small temporary drop in the energy confinement. In contrast, for both a very high confinement mode (VH-mode) and an H-mode with a broad internal transport barrier (ITB) extending over the entire core and coalesced with the edge transport barrier, the linearly unstable modes penetrate well into the mid radius and the corresponding consequences for global confinement are significantly more severe. The ELM accordingly results in an irreversible loss of the high performance

  3. Visible spectroscopy in the DIII-D divertor

    International Nuclear Information System (INIS)

    Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.; Tugarinov, S.; Whyte, D.G.

    1996-06-01

    Spectroscopy measurements in the DIII-D divertor have been carried out with a survey spectrometer which provides simultaneous registration of the visible spectrum over the region 400--900 nm with a resolution of 0.2 nm. Broad spectral coverage is achieved through use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges

  4. Multivariable shape control development on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    In this paper, the authors describe recent work on plasma shape and position control at DIII-D. This control consists of two equally challenging problems--the problem of identifying what the plasma actually looks like in real time, i.e. measuring the parameters to be controlled, and the task of determining the feedback algorithm which best controls these plasma parameters in a multiple input-output system. Recent implementation of the EFIT plasma equilibrium reconstruction algorithm in real time code which produces a new equilibrium estimate every 1.5 ms seems to solve the longstanding problem of obtaining sufficiently accurate plasma shape and position estimation. Stabilization of the open-loop unstable vertical motion is also viewed as a solved problem. The primary remaining problem appears to be how best to command the power supplies to achieve a desired shaping control response. They will describe the effort to understand and apply linearized models of plasma evolution to development and implementation of multivariable plasma controllers

  5. Characterization of wall conditions in DIII-D

    International Nuclear Information System (INIS)

    Holtrop, K.L.; Jackson, G.L.; Kellman, A.G.; Lee, R.L.; West, W.P.; Wood, R.D.; Whyte, D.G.

    1996-10-01

    Wall conditioning in DIII-D is one of the most important factors in achieving reproducible high confinement discharges. For example, the very high confinement mode (VH-mode) was only discovered after boronization, a CVD technique to deposit a thin boron film over the entire surface of the tokamak. In order to evaluate wall conditions and provide a data base to correlate these wall conditions with tokamak discharge performance, a series of nominally identical reference VH-mode discharges (1.6 MA, 2.1 T, double-null diverted) were taken at various times during a series of experimental campaigns with evolving wall conditions. These reference discharges have allowed a quantitative determination of how the wall conditions have evolved. For instance, core carbon and oxygen levels in the VH-mode phase remains at historically low levels during the 1995 run year and there was also a steady decrease in the oxygen levels at plasma initiation during this period. The authors discuss the long term changes in low Z impurities and the effect of wall conditioning techniques such as boronization and baking on these impurities. In addition, the evolution of the deuterium recycling rates will be discussed

  6. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Porter, G.D.; Wood, R.D.; Allen, S.L.; Boedo, J.; Brooks, N.H.; Evans, T.E.; Fenstermacher, M.E.; Hill, D.N.; Isler, R.C.; Lasnier, C.J.; Lehmer, R.D.; Mahdavi, M.A.; Maingi, R.; Moyer, R.A.; Petrie, T.W.; Schaffer, M.J.; Wade, M.R.; Watkins, J.G.; West, W.P.; Whyte, D.G.

    1998-01-01

    The radiation of divertor heat flux on DIII-D [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low-Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction-dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE [T. Rognlien, J. L. Milovich, M. E. Rensink, and G. D. Porter, J. Nucl. Mater. 196 endash 198, 347 (1992)] has reproduced many of the observed experimental features. copyright 1998 American Institute of Physics

  7. Grating spectrometer installation for electron cyclotron emission measurements on the DIII-D tokamak using circular waveguide and synchronous detection

    International Nuclear Information System (INIS)

    Lohr, J.; Jahns, G.; Moeller, C.; Prater, R.

    1986-01-01

    The grating spectrometer installation on the DIII-D tokamak uses fundamental circular waveguide propagating the TE 11 lowest-order mode followed by oversized circular guide carrying the low-loss TE 01 mode. The short section of fundamental guide permits use of an electronic chopper operating at 100 kHz for both calibration and plasma operation. By using ac-coupled amplifiers tuned to the chopping frequency, the background signal generated in the indium antimonide detectors by neutrons and x rays is automatically subtracted and the system noise bandwidth is reduced. Compared with a quasi-optical system, the much smaller fundamental horn and front-end waveguide allow the waveguide system to be located outside a gate valve. With this configuration the entire waveguide run, including the actual horn and vacuum window used during plasma operations, can be included in the calibration setup

  8. Grating spectrometer installation for electron cyclotron emission measurements on the DIII-D tokamak using circular waveguide and synchronous detection

    International Nuclear Information System (INIS)

    Lohr, J.; Jahns, G.; Moeller, C.; Prater, R.

    1986-03-01

    The grating spectrometer installation on the DIII-D tokamak uses fundamental circular waveguide propagating the TE 11 lowest order mode followed by oversized circular guide carrying the low loss TE 01 mode. The short section of fundamental guide permits use of an electronic chopper operating at 100 kHz for both calibration and plasma operation. By using ac-coupled amplifiers tuned to the chopping frequency, the background signal generated in the indium antimonide detectors by neutrons and x-rays is automatically subtracted and the system noise bandwidth is reduced. Compared with a quasi-optical system, the much smaller fundamental horn and front end waveguide allow the waveguide system to be located outside a gate valve. With this configuration the entire waveguide run, including the actual horn and vacuum window used during plasma operations, can be included in the calibration set-up

  9. CONTROL SYSTEM FOR THE LITHIUM BEAM EDGE PLASMA CURRENT DENSITY DIAGNOSTIC ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PEAVY, J.J.; CARY, W.P; THOMAS, D.M; KELLMAN, D.H.; HOYT, D.M; DELAWARE, S.W.; PRONKO, S.G.E.; HARRIS, T.E.

    2004-03-01

    OAK-B135 An edge plasma current density diagnostic employing a neutralized lithium ion beam system has been installed on the DIII-D tokamak. The lithium beam control system is designed around a GE Fanuc 90-30 series PLC and Cimplicity(reg s ign) HMI (Human Machine Interface) software. The control system operates and supervises a collection of commercial and in-house designed high voltage power supplies for beam acceleration and focusing, filament and bias power supplies for ion creation, neutralization, vacuum, triggering, and safety interlocks. This paper provides an overview of the control system, while highlighting innovative aspects including its remote operation, pulsed source heating and pulsed neutralizer heating, optimizing beam regulation, and beam ramping, ending with a discussion of its performance

  10. The vacuum vessel for the FTU device: design constraints and stress analysis

    International Nuclear Information System (INIS)

    Andreani, R.; Cecchini, A.; Gasparotto, M.; Lovisetto, L.; Migliori, S.; Pizzuto, A.

    1984-01-01

    The FTU vacuum vessel must withstand large electromagnetic loads due to the interactions between the eddy currents in the vessel and high magnetic fields of the machine, the atmospheric pressure and the severe thermal loads due to plasma losses and RF power not coupled to the plasma. In order to minimise the stresses on the vacuum chamber, an optimization of the wall thickness has been performed and, in order to assess the feasibility of the vessel, an extensive three dimensional finite element stress analysis has been developed. The main results obtained are illustrated. (author)

  11. Eddy currents in a nonperiodic vacuum vessel induced by axisymmetric plasma motion

    International Nuclear Information System (INIS)

    DeLucia, J.

    1985-12-01

    A method is described for calculating the two-dimensional trajectory of a vertically or horizontally unstable axisymmetric tokamak plasma in the presence of a resistive vacuum vessel. The vessel is not assumed to have toroidal symmetry. The plasma is represented by a current-filament loop that is free to move vertically and to change its major radius. Its position is evolved in time self-consistently with the vacuum vessel eddy currents. The plasma current, internal inductance, and poloidal beta can be specified functions of time so that eddy currents resulting from a disruption can be modeled. The vacuum vessel is represented by a set of current-filaments whose positions and orientations are chosen to model the dominant eddy current paths. Although the specific application is to TFTR, the present model is of general applicability. 7 refs., 4 figs., 2 tabs

  12. Electromagnetic forces on a metallic Tokamak vacuum vessel following a disruptive instability

    International Nuclear Information System (INIS)

    Eckhartt, D.

    1979-04-01

    During a 'hard' disruptive instability of a Tokamak plasma the current-carrying plasma is lost within a very short time, typically few milliseconds. If the plasma is contained in a metallic vacuum vessel, electric currents are set up in the vessel following the disappearance of the plasma current. These vessel currents together with the magnetic fields intersecting the vessel generate electromagnetic forces which appear as mechanical loads on the vessel. In the following note it is assumed that the vacuum vessel is surrounded by an 'outer equivalent' or 'flux-conserving' shell having a characteristic time of magnetic field penetration which is long compared to the time of existence of the vessel currents. This property defines the distribution of vessel current densities (and hence the load distribution) without referring to the exact mechanism or time sequence of events by which the plasma current is lost. Numerical examples of the electromagnetic force distribution from this model refer to parameters of the JET-device with the simplifying assumption of circular cross-sections for plasma current, vacuum vessel, and outer equivalent shell. (orig.)

  13. DIII-D research operations. Annual report, October 1, 1992--September 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    La Haye, R.J. [ed.

    1994-05-01

    The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R&D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma.

  14. Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

    2014-10-15

    Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

  15. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    International Nuclear Information System (INIS)

    Mozzillo, Rocco; Tarallo, Andrea; Marzullo, Domenico; Bachmann, Christian; Di Gironimo, Giuseppe; Mazzone, Giuseppe

    2016-01-01

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  16. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    Energy Technology Data Exchange (ETDEWEB)

    Mozzillo, Rocco, E-mail: rocco.mozzillo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Tarallo, Andrea; Marzullo, Domenico [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione - ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

    2016-11-15

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  17. Divertor plasma studies on DIII-D: Experiment and modeling

    International Nuclear Information System (INIS)

    West, W.P.; Brooks, N.H.; Allen, S.L.

    1996-09-01

    In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process

  18. Effects of neutrals on plasma rotation in DIII-D

    International Nuclear Information System (INIS)

    Monier-Garbet, P.; Burrell, K.H.; Hinton, F.L.; Kim, J.; Garbet, X.; Groebner, R.J.

    1997-01-01

    Friction due to charge exchange with cold neutral atoms in the edge is investigated as a candidate to govern the poloidal rotation in the edge of a tokamak plasma. The Hirshman and Sigmar neoclassical moment approach is used to determine the rotation velocities of the main plasma ions and of one impurity species, when charge exchange friction is included. It is found that the poloidal rotation of the main plasma ions is controlled by charge exchange friction with neutrals. The impurity ion poloidal rotation is governed by the balance between the impurity viscous force and the main-ion-impurity-ion friction force. The results of the calculation are compared with the measurements obtained in the edge of a DIII-D high (H) mode plasma, using charge exchange recombination (CER) spectroscopy. It is found that the measured main ion poloidal rotation can be accurately predicted by the neoclassical theory including the effect of neutrals, assuming a neutral density n > = 3 x 10 17 m -3 at the separatrix, decreasing exponentially inside the plasma with an e-folding length of 0.012 m, and peaking near the X point region with a poloidal peaking parameter y ≡ n > 2 >/ n B 2 > = 1.5. However, for the impurity ions, the neoclassical theory including a single impurity charge state, and regardless of the effect of the neutrals, gives a prediction that has the correct sign, but whose value is a factor of 5 or 6 different from the experimental value. (author). 12 refs, 7 figs, 1 tab

  19. INTEGATED ADVANCED TOKAMAK OPERATION ON DIII-D

    International Nuclear Information System (INIS)

    WADE, M.R.; MURAKAMI, M.; LUCE, T.C.; FERRON, J.R.; PETTY, C.C.; BRENNEN, D.P.; GAROFALO, A.M.; GREENFIELD, C.M.; HYATT, A.W.; JAYAKUMAR, R.; KINSEY, J.E.; La HAYE, R.J.; LAO, L.L.; LOHR, J.; POLITZER, P.A.; PRATER, R.; STRAIT, E.J.; WATKINS, J.G.

    2002-01-01

    Recent experiments on DIII-D have demonstrated the ability to sustain plasma conditions that integrate and sustain the key ingredients of Advanced Tokamak (AT) operation: high β with 1.5 min min > 2.0, plasmas with β ∼ 2.9% and 90% of the plasma current driven non-inductively have been sustained for nearly 2 s (limited only by the duration of the ECCD pulse). Negative central magnetic shear is produced by the ECCD, leading to the formation of a weak internal transport barrier even in the presence of Type I ELMs. Separate experiments have demonstrated the ability to sustain a steady current density profile using ECCD for periods as long as 1 s with β = 3.3% and > 90% of the current driven non-inductively. In addition, stable operation well above the ideal no-wall β limit has been sustained for several energy confinement times with the duration only limited by resistive relaxation of the current profile to an unstable state. Stability analysis indicates that the experimental β limit depends on the degree to which the no-wall limit can be exceeded and weakly on the actual no-wall limit. Achieving the necessary density levels required for adequate ECCD efficiency requires active divertor exhaust and reducing the wall inventory buildup prior to the high performance phase. Simulation studies indicate that the successful integration of high β operation with current profile control consistent with these experimental results should result in high β, fully non-inductive plasma operation

  20. Development of improved methods for remote access of DIII-D data and data analysis

    International Nuclear Information System (INIS)

    Greene, K.L.; McHarg, B.B. Jr.

    1997-11-01

    The DIII-D tokamak is a national fusion research facility. There is an increasing need to access data from remote sites in order to facilitate data analysis by collaborative researchers at remote locations, both nationally and internationally. In the past, this has usually been done by remotely logging into computers at the DIII-D site. With the advent of faster networking and powerful computers at remote sites, it is becoming possible to access and analyze data from anywhere in the world as if the remote user were actually at the DIII-D site. The general mechanism for accessing DIII-D data has always been via the PTDATA subroutine. Substantial enhancements are being made to that routine to make it more useful in a non-local environment. In particular, a caching mechanism is being built into PTDATA to make network data access more efficient. Studies are also being made of using Distributed File System (DFS) disk storage in a Distributed Computing Environment (DCE). A data server has been created that will migrate, on request, shot data from the DIII-D environment into the DFS environment

  1. Enhanced Computational Infrastructure for Data Analysis at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schissel, D.P.; Peng, Q.; Schachter, J.; Terpstra, T.B.; Casper, T.A.; Freeman, J.; Jong, R.; Keith, K.M.; Meyer, W.H.; Parker, C.T.; McCharg, B.B.

    1999-01-01

    Recently a number of enhancements to the computer hardware infrastructure have been implemented at the DIII-D National Fusion Facility. Utilizing these improvements to the hardware infrastructure, software enhancements are focusing on streamlined analysis, automation, and graphical user interface (GUI) systems to enlarge the user base. The adoption of the load balancing software package LSF Suite by Platform Computing has dramatically increased the availability of CPU cycles and the efficiency of their use. Streamlined analysis has been aided by the adoption of the MDSplus system to provide a unified interface to analyzed DIII-D data. The majority of MDSplus data is made available in between pulses giving the researcher critical information before setting up the next pulse. Work on data viewing and analysis tools focuses on efficient GUI design with object-oriented programming (OOP) for maximum code flexibility. Work to enhance the computational infrastructure at DIII-D has included a significant effort to aid the remote collaborator since the DIII-D National Team consists of scientists from 9 national laboratories, 19 foreign laboratories, 16 universities, and 5 industrial partnerships. As a result of this work, DIII-D data is available on a 24 x 7 basis from a set of viewing and analysis tools that can be run either on the collaborators' or DIII-Ds computer systems. Additionally, a Web based data and code documentation system has been created to aid the novice and expert user alike

  2. Helium Exhaust Studies in H-Mode Discharges in the DIII-D Tokamak Using an Argon-Frosted Divertor Cryopump

    International Nuclear Information System (INIS)

    Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P.; Brooks, N.H.; Burrell, K.H.; Groebner, R.J.; Jackson, G.L.; Klepper, C.C.; Laughon, G.; Menon, M.M.; Mioduszewski, P.K.

    1995-01-01

    The first experiments demonstrating exhaust of thermal helium in a diverted, H-mode deuterium plasma have been performed on the DIII-D tokamak. The helium, introduced via gas puffing, is observed to reach the plasma core, and then is readily removed from the plasma with a time constant of ∼10--20 energy-confinement times by an in-vessel cryopump conditioned with argon frosting. Detailed analysis of the helium profile evolution suggests that the exhaust rate is limited by the exhaust efficiency of the pump (∼5%) and not by the intrinsic helium-transport properties of the plasma

  3. Design of the Intersector Welding Robot for vacuum vessel assembly and maintenance

    International Nuclear Information System (INIS)

    Jones, L.; Dagenais, J.-F.; Daenner, W.; Maisonnier, D.

    2000-01-01

    Next Step Fusion Devices require on-site (field weld) joining of sectors of the thick-walled vacuum vessel for structural and vacuum integrity. EFDA (European Fusion Development Agreement) is supporting an R and D programme to investigate processes for assembly of the vacuum vessel and to carry out cutting, re-welding and inspection for remote sector replacement, forming part of the overall VV/blanket research effort. In order to direct the process end-effectors along the field joint zone, a track-mounted Intersector Welding Robot (IWR) on a mock-up of a region of the vacuum vessel has been designed and is described in this paper. A rail-mounted hexapod type robot offers six axes of motion over a limited work envelope with high payload to robot weight ratio. A solution to the production of reduced pressure local vacuum is the installation of short, lightweight segments bolted to each other and the vessel wall. The various process heads can be mounted using end-effectors of special design. To minimise the supply and interface problems for the IWR prototype, its motion control and electronic systems will be embedded locally. A laser scan with camera forms the on-line seam tracking capability to compensate for rail and seam deviations

  4. Evaluation of structural reliability for vacuum vessel under external pressure and electromagnetic force

    International Nuclear Information System (INIS)

    Minato, Akio

    1983-08-01

    Static and dynamic structural analyses of the vacuum vessel for a Swimming Pool Type Tokamak Reactor (SPTR) have been conducted under the external pressure (hydraulic and atmospheric pressure) during normal operation or the electromagnetic force due to plasma disruption. The reactor structural design is based on the concept that the adjacent modules of the vacuum vessel are not connected mechanically with bolts in the torus inboard region each other, so as to save the required space for inserting the remote handling machine for tightenning and untightenning bolts in the region and to simplify the repair and maintenance of the reactor. The structural analyses of the vacuum vessel have been carried out under the external pressure and the electromagnetic force and the structural reliability against the static and dynamic loads is estimated. The several configurations of the lip seal between the modules, which is required to make a plasma vacuum boundary, have been proposed and the structural strength under the forced displacements due to the deformation of the vacuum vessel is also estimated. (author)

  5. Fabrication of the vacuum vessel of the Spanish stellarator TJ-II

    International Nuclear Information System (INIS)

    Botija, J.; Blaumoser, M.; Cal, E. de la; Garcia, A.; Tabares, F.; Molleta, L.; Rigadello, D.; Dal Maso, S.; Bevilacqua, G.

    1995-01-01

    TJ-II is a medium size stellarator under construction in Madrid, Spain. Its major plasma radius is 1.5 m and its minor plasma dimensions are 0.2m by 0.4m. The toroidal magnetic field on the axis is 1T. The bean shaped helical plasma is contained in a stainless steel vacuum vessel with a total of 96 ports, including 8 manholes to have access to its interior. The vacuum vessel will be baked at 150 C. Its complicated geometry along with the high tolerance requirements make this component a difficult manufacturing challenge. (orig.)

  6. Integration of cooking and vacuum cooling of carrots in a same vessel

    Directory of Open Access Journals (Sweden)

    Luiz Gustavo Gonçalves Rodrigues

    2012-03-01

    Full Text Available Cooked vegetables are commonly used in the preparation of ready-to-eat foods. The integration of cooking and cooling of carrots and vacuum cooling in a single vessel is described in this paper. The combination of different methods of cooking and vacuum cooling was investigated. Integrated processes of cooking and vacuum cooling in a same vessel enabled obtaining cooked and cooled carrots at the final temperature of 10 ºC, which is adequate for preparing ready-to-eat foods safely. When cooking and cooling steps were performed with the samples immersed in boiling water, the effective weight loss was approximately 3.6%. When the cooking step was performed with the samples in boiling water or steamed, and the vacuum cooling was applied after draining the boiling water, water loss ranged between 15 and 20%, which caused changes in the product texture. This problem can be solved with rehydration using a small amount of sterile cold water. The instrumental textural properties of carrots samples rehydrated at both vacuum and atmospheric conditions were very similar. Therefore, the integrated process of cooking and vacuum cooling of carrots in a single vessel is a feasible alternative for processing such kind of foods.

  7. Structural analysis and manufacture for the vacuum vessel of experimental advanced superconducting tokamak (EAST) device

    International Nuclear Information System (INIS)

    Song Yuntao; Yao Damao; Wu Songata; Weng Peide

    2006-01-01

    The experimental advanced superconducting tokamak (EAST) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and is being constructed as the Chinese national nuclear fusion research project. The vacuum vessel, that is one of the key components, will have to withstand not only the electromagnetic force due to the plasma disruption and the Halo current, but also the pressure of boride water and the thermal stress due to the 250 deg. C baking out by the hot pressure nitrogen gas, or the 100 deg. C hot wall during plasma operation. This paper is a report of the mechanical analyses of the vacuum vessel. According to the allowable stress criteria of American Society of Mechanical Engineers, Boiler and Pressure Vessel Committee (ASME), the maximum integrated stress intensity on the vacuum vessel is 396 MPa, less than the allowable design stress intensity 3S m (441 MPa). At the same time, some key R and D issues are presented, which include supporting system, bellows and the assembly of the whole vacuum vessel

  8. Design, fabrication and test of double-wall vacuum vessel for JT-60U

    International Nuclear Information System (INIS)

    Uchikawa, Takashi; Ioki, Kimihiro; Ninomiya, Hiromasa.

    1994-01-01

    A double-wall vacuum vessel was designed and fabricated for JT-60U (an upgraded machine of JT-60), which has a plasma current up to 6 MA and a large plasma volume (100 m 3 ). A new concept of Inconel 625 all-welded structure was adopted to the vessel, that comprises an inner plate, square tubes and an outer plate. The vacuum vessel with a multi-arc D-shaped cross section was fabricated by using hot-sizing press. The electromagnetic and structural analysis has been performed for plasma disruption loads. Dynamic responses of the vessel were measured during plasma disruptions, and the observed displacement had a good agreement with the result of FEM analysis. (author)

  9. ITER vacuum vessel design (D201 subtask 1.3 and subtask 3). Final report

    International Nuclear Information System (INIS)

    1996-01-01

    ITER Task No. D201, Vacuum Vessel Design (Subtask 1.3 and Subtask 3), was initiated to propose and evaluate local vacuum vessel reinforcement alternatives in proximity to the Neutral Beam, Radial Mid-Plane, Top, and Divertor Ports. These areas were reported to be highly stressed regions based on the results of preliminary stress analyses performed by the USHT (US Home Team) and the ITER Joint Central Team (JCT) at the Garching JWS (Joint Work Site). Initial design activities focused on the divertor port region which was reported to experience the highest stress intensities. Existing stress analysis models and results were reviewed with the USHT stress analysts to obtain an overall understanding of the vessel response to the various applied loads. These reviews indicated that the reported stress intensities in the divertor port region were significantly affected by the loads applied to the vessel in adjacent regions

  10. Neutronics studies for the design of the European DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

    2016-11-01

    Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

  11. ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

    International Nuclear Information System (INIS)

    HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

    2002-01-01

    A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

  12. Extending the capabilities of the DIII-D Plasma Control System for worldwide fusion research collaborations

    International Nuclear Information System (INIS)

    Penaflor, B.G.; Ferron, J.R.; Walker, M.L.; Humphreys, D.A.; Leuer, J.A.; Piglowski, D.A.; Johnson, R.D.; Xiao, B.J.; Hahn, S.H.; Gates, D.A.

    2009-01-01

    This paper will discuss the recent enhancements which have been made to the DIII-D Plasma Control System (PCS) in order to further extend its usefulness as a shared tool for worldwide fusion research. The PCS developed at General Atomics is currently being used in a number of fusion research experiments worldwide, including the DIII-D Tokamak Facility in San Diego, and most recently the KSTAR Tokamak in South Korea. A number of enhancements have been made to support the ongoing needs of the DIII-D Tokamak in addition to meeting the needs of other PCS users worldwide. Details of the present PCS hardware and software architecture along with descriptions of the latest enhancements will be given.

  13. Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-04-01

    Test specimens made with the 832665 heat of V-4Cr-4Ti alloy were exposed in the DIII-D tokamak environment to support the installation of components made of a V-4Cr-4Ti alloy in the radiative divertor of the DIII-D. Some of the tests were conducted with the Divertor Materials Evaluation System (DiMES) to study the short-term effects of postvent bakeout, when concentrations of gaseous impurities in the DIII-D chamber are the highest. Other specimens were mounted next to the chamber wall behind the divertor baffle plate, to study the effects of longer-term exposures. By design, none of the specimens directly interacted with the plasma. Preliminary results from testing the exposed specimens indicate only minor degradation of mechanical properties. Additional testing and microstructural characterization are in progress.

  14. Performance, diagnostics, controls and plans for the gyrotron system on the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    Ponce D.M.

    2012-09-01

    Full Text Available The DIII-D ECH complex is being upgraded with three new depressed collector gyrotrons. The performance of the existing system has been very good. As more gyrotrons having higher power are added to the system, diagnostics of gyrotron operation, optimization of the performance and qualification of components for higher power become more important. A new FPGA-based gyrotron control system is being installed, additional capabilities for rapid real time variation of the rf injection angles by the DIII-D Plasma Control System are being tested and infrastructure enhancements are being completed. Longer term plans continue to include ECH as a major component in the DIII-D heating and current drive capabilities.

  15. Increasing the Tokamak Pressure Limit: Tearing Mode Experiments in DIII-D

    International Nuclear Information System (INIS)

    La Haye, R.J.

    2005-01-01

    Since its reconfiguration in 1986, DIII-D has performed a number of experiments involving resistive magnetohydrodynamic (MHD) stability. These were and are directed to understand the conditions in which confinement and beta reducing tearing mode islands form, how to avoid them, and if unavoidable, how to stabilize them. Coils for correction of toroidal nonaxisymmetry have been developed to avoid error field locked mode islands. Basic classical tearing mode stability physics has been confirmed with a state-of-the-art ensemble of profile diagnostics, MHD equilibrium reconstruction, and stability code analysis. Neoclassical tearing mode thresholds and seeding are now much better understood with future large higher field devices expected to be 'metastable'. DIII-D is the leader in sophisticated real-time alignment of stabilizing electron cyclotron current drive on otherwise unstable rational surfaces. In all, DIII-D experiments are showing how higher stable beta with good confinement can be maintained without tearing mode islands limiting the plasma pressure

  16. Divertor heat and particle control experiments on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Baker, D.R.; Allen, S.L.

    1994-05-01

    In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D 2 gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models

  17. Regime of very high confinement in the boronized DIII-D tokamak

    International Nuclear Information System (INIS)

    Jackson, G.L.; Winter, J.; Taylor, T.S.; Burrell, K.H.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Hodapp, T.; Holtrop, K.; Lazarus, E.A.; Lao, L.L.; Lippmann, S.I.; Osborne, T.H.; Petrie, T.W.; Phillips, J.; James, R.; Schissel, D.P.; Strait, E.J.; Turnbull, A.D.; West, W.P.; DIII-D Team

    1991-01-01

    Following boronization, tokamak discharges in DIII-D have been obtained with confinement times up to a factor of 3.5 above the ITER89-P L-mode scaling and 1.8 times greater than the DIII-D/JET H-mode scaling relation. Very high confinement phases are characterized by relatively high central density with n e (0)∼1x10 20 m -3 , and central ion temperatures up to 13.6 keV at moderate plasma currents (1.6 MA) and heating powers (12.5--15.3 MW). These discharges exhibit a low fraction of radiated power, P≤25%, Z eff (0) close to unity, and lower impurity influxes than comparable DIII-D discharges before boronization

  18. GYRO Simulations of Core Momentum Transport in DIII-D and JET Plasmas

    International Nuclear Information System (INIS)

    Budny, R.V.; Candy, J.; Waltz, R.E.

    2005-01-01

    Momentum, energy, and particle transport in DIII-D and JET ELMy H-mode plasmas is simulated with GYRO and compared with measurements analyzed using TRANSP. The simulated transport depends sensitively on the nabla(T(sub)i) turbulence drive and the nabla(E(sub)r) turbulence suppression inputs. With their nominal values indicated by measurements, the simulations over-predict the momentum and energy transport in the DIII-D plasmas, and under-predict in the JET plasmas. Reducing |nabla(T(sub)i)| and increasing |nabla(E(sub)r)| by up to 15% leads to approximate agreement (within a factor of two) for the DIII-D cases. For the JET cases, increasing |nabla(T(sub)i)| or reducing |nabla(E(sub)r)| results in approximate agreement for the energy flow, but the ratio of the simulated energy and momentum flows remains higher than measurements by a factor of 2-4

  19. Temperature field and thermal stress analysis of the HT-7U vacuum vessel

    International Nuclear Information System (INIS)

    Song Yuntao; Yao Damao; Wu Songtao; Weng Peide

    2000-01-01

    The HT-7U vacuum vessel is an all-metal-welded double-wall interconnected with toroidal and poloidal stiffening ribs. The channels formed between the ribs and walls are filled with boride water as a nuclear shielding. On the vessel surface facing the plasma are installed cable-based Ohmic heaters. Prior to plasma operation the vessel is to be baked out and discharge cleaned at about 250 degree C. During baking out the non-uniformity of temperature distribution on the vacuum vessel will bring about serious thermal stress that can damage the vessel. In order to determine and optimize the design of the HT-7U vacuum vessel, a three-dimensional finite element model was performed to analyse its temperature field and thermal stress. the maximal thermal stress appeared on the round of lower vertical port and maximal deformation located just on the region between the upper vertical port and the horizontal port. The results show that the reinforced structure has a good capability of withstanding the thermal loads

  20. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    Science.gov (United States)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  1. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    International Nuclear Information System (INIS)

    Kumar, B Ramesh; Gangradey, R

    2012-01-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  2. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1990--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Simonen, T.C.; Evans, T.E. (eds.)

    1992-03-01

    This report discusses the following topics on Doublet-3 research operations: DIII-D Program Overview; Boundary Plasma Research Program/Scientific Progress; Radio Frequency Heating and Current Drive; Core Physics; DIII-D Operations; Program Development; Support Services; ITER Contributions; Burning Plasma Experiment Contributions; and Collaborative Efforts.

  3. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1990--September 30, 1991

    International Nuclear Information System (INIS)

    Simonen, T.C.; Evans, T.E.

    1992-03-01

    This report discusses the following topics on Doublet-3 research operations: DIII-D Program Overview; Boundary Plasma Research Program/Scientific Progress; Radio Frequency Heating and Current Drive; Core Physics; DIII-D Operations; Program Development; Support Services; ITER Contributions; Burning Plasma Experiment Contributions; and Collaborative Efforts

  4. Thermal structural analysis of SST-1 vacuum vessel and cryostat assembly using ANSYS

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Bedakihale, Vijay; Ranganath, Tata

    2009-01-01

    Steady state super-conducting tokamak-1 (SST-1) is a medium sized tokamak, which has been designed to produce a 'D' shaped double null divertor plasma and operate in quasi steady state (1000 s). SST-1 vacuum system comprises of plasma chamber (vacuum vessel, interconnecting rings, baking and cooling channels), and cryostat all made of SS 304L material designed to meet ultra high vacuum requirements for plasma generation and confinement. Prior to plasma shot and operation the vessel assembly is baked to 250/150 deg. C from room temperature and discharge cleaned to remove impurities/trapped gases from wall surfaces. Due to baking the non-uniform temperature pattern on the vessel assembly coupled with atmospheric pressure loading and self-weight give rise to high thermal-structural stresses, which needs to be analyzed in detail. In addition the vessel assembly being a thin shell vessel structure needs to be checked for critical buckling load caused by atmospheric and baking thermal loads. Considering symmetry of SST-1, 1/16th of the geometry is modeled for finite element (FE) analysis using ANSYS for different loading scenarios, e.g. self-weight, pressure loading considering normal operating conditions, and off-normal loads coupled with baking of vacuum vessel from room temperature 250 deg. C to 150 deg. C, buckling and modal analysis for future dynamic analysis. The paper will discuss details about SST-1 vacuum system/cryostat, solid and FE model of SST-1, different loading scenarios, material details and the stress codes used. We will also present the thermal structural results of FE analysis using ANSYS for various load cases being investigated and our observations under different loading conditions.

  5. ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY

    International Nuclear Information System (INIS)

    PRATER, R; PETTY, CC; LUCE, TC; HARVEY, RW; CHOI, M; LAHAYE, RJ; LIN-LIU, Y-R; LOHR, J; MURAKAMI, M; WADE, MR; WONG, K-L

    2003-01-01

    A271 ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY. Experiments on the DIII-D tokamak in which the measured off-axis electron cyclotron current drive has been compared systematically to theory over a broad range of parameters have shown that the Fokker-Planck code CQL3D provides an excellent model of the relevant current drive physics. This physics understanding has been critical in optimizing the application of ECCD to high performance discharges, supporting such applications as suppression of neoclassical tearing modes and control and sustainment of the current profile

  6. Analysis of toroidal rotation data for the DIII-D tokamak

    International Nuclear Information System (INIS)

    St John, H.; Stroth, U.; Burrell, K.H.; Groebner, R.J.; DeBoo, J.C.; Gohil, P.

    1989-01-01

    Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Our results are based on numerical inversions using the transport code ONETWO, modified to account for the radial diffusion of toroidal angular momentum. 13 refs., 4 figs

  7. Analysis of toroidal rotation data for the DIII-D tokamak

    International Nuclear Information System (INIS)

    John, H.St.; Burrell, K.H.; Groebner, R.; DeBoo, J.; Gohil, P.

    1989-01-01

    Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner et al. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Similar studies have been previously reported for Doublet III, ASDEX, TFTR, JET and other tokamaks. (author) 13 refs., 4 figs

  8. Development of a new error field correction coil (C-coil) for DIII-D

    International Nuclear Information System (INIS)

    Robinson, J.I.; Scoville, J.T.

    1995-12-01

    The C-coil recently installed on the DIII-D tokamak was developed to reduce the error fields created by imperfections in the location and geometry of the existing coils used to confine, heat, and shape the plasma. First results from C-coil experiments include stable operation in a 1.6 MA plasma with a density less than 1.0 x 10 13 cm -3 , nearly a factor of three lower density than that achievable without the C-coil. The C-coil has also been used in magnetic braking of the plasma rotation and high energy particle confinement experiments. The C-coil system consists of six individual saddle coils, each 60 degree wide toroidally, spanning the midplane of the vessel with a vertical height of 1.6 m. The coils are located at a major radius of 3.2 m, just outside of the toroidal field coils. The actual shape and geometry of each coil section varied somewhat from the nominal dimensions due to the large number of obstructions to the desired coil path around the already crowded tokamak. Each coil section consists of four turns of 750 MCM insulated copper cable banded with stainless steel straps within the web of a 3 in. x 3 in. stainless steel angle frame. The C-coil structure was designed to resist peak transient radial forces (up to 1,800 Nm) exerted on the coil by the toroidal and ploidal fields. The coil frames were supported from existing poloidal field coil case brackets, coil studs, and various other structures on the tokamak

  9. Development of remote welding equipment and techniques for the TFTR vacuum vessel

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Larson, R.A.; Aldrich, W.C.

    1980-01-01

    In the event that the TFTR vacuum vessel is damaged or one of the toroidal field coils fails after the system has become substantially activated, it is necessary to remotely remove and replace the damaged section of the vessel using remote handling procedures. This paper describes a welding system developed through the final design stage to perform the remote welding necessary during the replacement operation. Information is presented describing the vessel configuration, the replacement sequence, the welding system requirements, welder configuration, supporting systems, the weld development program and future development requirements

  10. Calculation of voltages and currents induced in the vacuum vessel of ASDEX by plasma disruptions

    International Nuclear Information System (INIS)

    Preis, H.

    1978-01-01

    An approximation method is used to analyze the electromagnetic diffusion process induced in the walls of the ASDEX vacuum vessel by plasma disruptions. For this purpose the rotational-symmetric vessel is regarded as N = 82 circular conductors connected in parallel and inductively coupled with one another and with the plasma. The transient currents and voltages occurring in this circuit are calculated with computer programs. From the calculated currents it is possible to determine the time behavior of the distributions of the current density and magnetic force density in the vessel walls. (orig.) [de

  11. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R. [ITER JCT, Garching (Germany)

    1998-10-01

    Design and R and D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R and D results. The resulting design changes are discussed for each system. (orig.) 11 refs.

  12. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Science.gov (United States)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R.

    1998-10-01

    Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

  13. Modeling the Thermal Mechanical Behavior of a 300 K Vacuum Vessel that is Cooled by Liquid Hydrogen in Film Boiling

    International Nuclear Information System (INIS)

    Yang, S.Q.; Green, M.A.; Lau, W.

    2004-01-01

    This report discusses the results from the rupture of a thin window that is part of a 20-liter liquid hydrogen vessel. This rupture will spill liquid hydrogen onto the walls and bottom of a 300 K cylindrical vacuum vessel. The spilled hydrogen goes into film boiling, which removes the thermal energy from the vacuum vessel wall. This report analyzes the transient heat transfer in the vessel and calculates the thermal deflection and stress that will result from the boiling liquid in contact with the vessel walls. This analysis was applied to aluminum and stainless steel vessels

  14. 13C-Tracer Experiments in DIII-D Preliminary to Thermal Oxidation Experiments to Understand Tritium Recovery in DIII-D, JET, C-Mod, and MAST

    International Nuclear Information System (INIS)

    Stangeby, P.; Allen, S.; Bekris, N.; Brooks, N.; Christie, K.; Chrobak, C.; Coad, J.; Counsell, G.; Davis, J.; Elder, J.; Fenstermacher, M.; Groth, M.; Haasz, A.; Likonen, J.; Lipschultz, B.; McLean, A.; Philipps, V.; Porter, G.; Rudakov, D.; Shea, J.; Wampler, W.; Watkins, J.; West, W.; Whyte, D.

    2006-01-01

    Retention of tritium in carbon co-deposits is a serious concern for ITER. Developing a reliable in-situ removal method of the co-deposited tritium would allow the use of carbon plasma-facing components which have proven reliable in high heat flux conditions and compatible with high performance plasmas. Thermal oxidation is a potential solution, capable of reaching even hidden locations. It is necessary to establish the least severe conditions to achieve adequate tritium recovery, minimizing damage and reconditioning time. The first step in this multi-machine project is 13 C-tracer experiments in DIII-D, JET, C-Mod and MAST. In DIII-D and JET, 13 CH 4 has been (and in C-Mod and MAST, will be) injected toroidally symmetrically, facilitating quantification and interpretation of the results. Tiles have been removed, analyzed for 13 C content and will next be evaluated in a thermal oxidation test facility in Toronto with regard to the ability of different severities of oxidation exposure to remove the different types of (known and measured) 13 C co-deposit. Removal of D/T from B on Mo tiles from C-Mod will also be tested. OEDGE interpretive code analysis of the 13 C deposition patterns is used to generate the understanding needed to apply findings to ITER. First results are reported here for the 13 C injection experiments IN DIII-D

  15. Design and fabrication of the vacuum vessel for the Advanced Toroidal Facility

    International Nuclear Information System (INIS)

    Chipley, K.K.; Frey, G.N.

    1985-01-01

    The vacuum vessel for the Advanced Toroidal Facility (ATF) is a heavily contoured and very complex formed vessel that is specifically designed to allow for maximum plasma volume in a pure stellarator arrangement. The design of the facility incorporates an internal vessel that is closely fitted to the two helical field coils following the winding law theta = 1/6phi. Metallic seals have been incorporated throughout the system to minimize impurities. The vessel has been fabricated utilizing a comprehensive set of tooling fixtures specifically designed for the task of forming 6-mm stainless steel plate to the complex shape. Computer programs were used to develop a series of ribs that essentially form an internal mold of the vessel. Plates were press-formed with multiple compound curves, fitted to the fixture, and joined with full-penetration welds. 7 refs., 8 figs

  16. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    International Nuclear Information System (INIS)

    Joshi, Jaydeep; Yadav, Ashish; Gangadharan, Roopesh; Prasad, Rambilas; Ulahannan, Shino; Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun

    2015-01-01

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  17. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jaydeep, E-mail: Jaydeep.joshi@iter-india.org [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Yadav, Ashish; Gangadharan, Roopesh [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Prasad, Rambilas [Madan Mohan Malaviya University of Technology, Gorakhpur, Uttar Pradesh 273001 (India); Ulahannan, Shino [Airframe Aerodesigns Pvt. Ltd., HAL Airport Exit Road, Old Airport Road, Bengaluru 17 (India); Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India)

    2015-10-15

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  18. General and crevice corrosion study of the in-wall shielding materials for ITER vacuum vessel

    Science.gov (United States)

    Joshi, K. S.; Pathak, H. A.; Dayal, R. K.; Bafna, V. K.; Kimihiro, Ioki; Barabash, V.

    2012-11-01

    Vacuum vessel In-Wall Shield (IWS) will be inserted between the inner and outer shells of the ITER vacuum vessel. The behaviour of IWS in the vacuum vessel especially concerning the susceptibility to crevice of shielding block assemblies could cause rapid and extensive corrosion attacks. Even galvanic corrosion may be due to different metals in same electrolyte. IWS blocks are not accessible until life of the machine after closing of vacuum vessel. Hence, it is necessary to study the susceptibility of IWS materials to general corrosion and crevice corrosion under operations of ITER vacuum vessel. Corrosion properties of IWS materials were studied by using (i) Immersion technique and (ii) Electro-chemical Polarization techniques. All the sample materials were subjected to a series of examinations before and after immersion test, like Loss/Gain weight measurement, SEM analysis, and Optical stereo microscopy, measurement of surface profile and hardness of materials. After immersion test, SS 304B4 and SS 304B7 showed slight weight gain which indicate oxide layer formation on the surface of coupons. The SS 430 material showed negligible weight loss which indicates mild general corrosion effect. On visual observation with SEM and Metallography, all material showed pitting corrosion attack. All sample materials were subjected to series of measurements like Open Circuit potential, Cyclic polarization, Pitting potential, protection potential, Critical anodic current and SEM examination. All materials show pitting loop in OC2 operating condition. However, its absence in OC1 operating condition clearly indicates the activity of chloride ion to penetrate oxide layer on the sample surface, at higher temperature. The critical pitting temperature of all samples remains between 100° and 200°C.

  19. Project management techniques used in the European Vacuum Vessel sectors procurement for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Losasso, Marcello, E-mail: marcello.losasso@f4e.europa.eu [Fusion for Energy (F4E), Barcelona (Spain); Ortiz de Zuniga, Maria; Jones, Lawrence; Bayon, Angel; Arbogast, Jean-Francois; Caixas, Joan; Fernandez, Jose; Galvan, Stefano; Jover, Teresa [Fusion for Energy (F4E), Barcelona (Spain); Ioki, Kimihiro [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lewczanin, Michal; Mico, Gonzalo; Pacheco, Jose Miguel [Fusion for Energy (F4E), Barcelona (Spain); Preble, Joseph [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Stamos, Vassilis; Trentea, Alexandru [Fusion for Energy (F4E), Barcelona (Spain)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer File name contains the directory tree structure with a string of three-letter acronyms, thereby enabling parent directory location when confronted with orphan files. Black-Right-Pointing-Pointer The management of the procurement procedure was carried out in an efficient and timely manner, achieving precisely the contract placement date foreseen at the start of the process. Black-Right-Pointing-Pointer The contract start-up has been effectively implemented and a flexible project management system has been put in place for an efficient monitoring of the contract. - Abstract: The contract for the seven European Sectors of the ITER Vacuum Vessel (VV) was placed at the end of 2010 with a consortium of three Italian companies. The task of placing and the initial take-off of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. A stringent quality controlled system with a bespoke Vacuum Vessel Project Lifecycle Management system to control the information flow, based on ENOVIA SmarTeam, was developed to handle the storage and approval of Documentation including links to the F4E Vacuum Vessel system and ITER International Organization System interfaces. The VV Sector design and manufacturing schedule is based on Primavera software, which is cost loaded thus allowing F4E to carry out performance measurement with respect to its payments and commitments. This schedule is then integrated into the overall Vacuum Vessel schedule, which includes ancillary activities such as instruments, preliminary design and analysis. The VV Sector Risk Management included three separate risk analyses from F4E and the bidders, utilizing two different methodologies. These efforts will lead to an efficient and effective implementation of this contract, vital to the success of the ITER machine, since the Vacuum Vessel is the biggest single work package of Europe's contribution to ITER and

  20. Electromagnetic loads and structural response of the CIT [Compact Ignition Tokamak] vacuum vessel to plasma disruptions

    International Nuclear Information System (INIS)

    Salem, S.L.; Listvinsky, G.; Lee, M.Y.; Bailey, C.

    1987-01-01

    Studies of the electromagnetic loads produced by a variety of plasma disruptions, and the resulting structural effects on the compact Ignition Tokamak (CIT) vacuum vessel (VV), have been performed to help optimize the VV design. A series of stationary and moving plasmas, with disruption rates from 0.7--10.0 MA/ms, have been analyzed using the EMPRES code to compute eddy currents and electromagnetic pressures, and the NASTRAN code to evaluate the structural response of the vacuum vessel. Key factors contributing to the magnitude of EM forces and resulting stresses on the vessel have been found to include disruption rate, and direction and synchronization of plasma motion with the onset of plasma current decay. As a result of these analyses, a number of design changes have been made, and design margins for the present 1.75 meter design have been improved over the original CIT configuration. 1 ref., 10 figs., 4 tabs

  1. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W.R.; Smith, J.P.; Trester, P.W.

    1997-04-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy.

  2. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    International Nuclear Information System (INIS)

    Johnson, W.R.; Smith, J.P.; Trester, P.W.

    1997-01-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy

  3. Advances in the operation of the DIII-D neutral beam computer systems

    International Nuclear Information System (INIS)

    Phillips, J.C.; Busath, J.L.; Penaflor, B.G.; Piglowski, D.; Kellman, D.H.; Chiu, H.K.; Hong, R.M.

    1998-02-01

    The DIII-D neutral beam system routinely provides up to 20 MW of deuterium neutral beam heating in support of experiments on the DIII-D tokamak, and is a critical part of the DIII-D physics experimental program. The four computer systems previously used to control neutral beam operation and data acquisition were designed and implemented in the late 1970's and used on DIII and DIII-D from 1981--1996. By comparison to modern standards, they had become expensive to maintain, slow and cumbersome, making it difficult to implement improvements. Most critical of all, they were not networked computers. During the 1997 experimental campaign, these systems were replaced with new Unix compliant hardware and, for the most part, commercially available software. This paper describes operational experience with the new neutral beam computer systems, and new advances made possible by using features not previously available. These include retention and access to historical data, an asynchronously fired ''rules'' base, and a relatively straightforward programming interface. Methods and principles for extending the availability of data beyond the scope of the operator consoles will be discussed

  4. 60 MHz fast wave current drive experiment for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Mayberry, M.J.; Chiu, S.C.; Porkolab, M.; Chan, V.; Freeman, R.; Harvey, R.; Pinsker, R. (General Atomics, San Diego, CA (USA))

    1989-07-01

    The DIII-D facility provides an opportunity to test fast wave current drive appoach. Efficient FWCD is achieved by direct electron absorption due to Landa damping and transit time magnetic pumping. To avoid competing damping mechamisms we seek to maximize the single-pass asorption of the fast waves by electrons. (AIP)

  5. Cooperative program to analyze heat and particle transport at high beta in DIII-D

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1990-01-01

    The objective is to collaborate with the General Atomics staff and the LLNL staff at General Atomics in the analysis of transport data from DIII-D. The Berkeley effort is integrated into the ongoing efforts at GA to help expedite progress in the fundamental understanding of transport phenomena in tokamaks

  6. Physics of turbulence control and transport barrier formation in DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; Burrell, K.H.; Carlstrom, T.N.

    1996-10-01

    This paper describes the physical mechanisms responsible for turbulence control and transport barrier formation on DIII-D as determined from a synthesis of results from different enhanced confinement regimes, including quantitative and qualitative comparisons to theory. A wide range of DIII-D data support the hypothesis that a single underlying physical mechanism, turbulence suppression via E x B shear flow is playing an essential, though not necessarily unique, role in reducing turbulence and transport in all of the following improved confinement regimes: H-mode, VH-mode, high-ell i modes, improved performance counter-injection L-mode discharges and high performance negative central shear (NCS) discharges. DIII-D data also indicate that synergistic effects are important in some cases, as in NCS discharges where negative magnetic shear also plays a role in transport barrier formation. This work indicates that in order to control turbulence and transport it is important to focus on understanding physical mechanisms, such as E x B shear, which can regulate and control entire classes of turbulent modes, and thus control transport. In the highest performance DIII-D discharges, NCS plasmas with a VH-mode like edge, turbulence is suppressed at all radii, resulting in neoclassical levels of ion transport over most of the plasma volume

  7. Data analysis software tools for enhanced collaboration at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schachter, J.; Peng, Q.; Schissel, D.P.

    2000-01-01

    Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is 'GAPlotObj', an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is 'MDSplus', which is used at DIII-D as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are 'ReviewPlus' and 'EFITviewer'. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server

  8. Comparison of calculated neutral beam shine through with measured shine-through in DIII-D

    International Nuclear Information System (INIS)

    Chiu, H.K.; Hong, R.

    1997-11-01

    A comparison of the calculated shine through of neutral particle beams in the DIII-D plasma to measured values inferred from the target temperature rise is reported. This provides an opportunity to verify the shine through calculations and makes them more reliable in those cases where the shine through can not be measured. The DIII-D centerpost neutral beam target tiles are safe-guarded against excessive beam shine-through by pyrometry and thermocouple (TC) arrays on the tiles. Shine-through beam power is calculated from the measured temperature changes reported by the target tile TC array. These measurements are performed at the beginning of each operational year at DIII-D. Theoretically, the beam energy deposited into the plasma can be expressed as a function of the change in beam density. Neutral beam energy deposition in plasma (of known density) is inferred by comparing the results of a series of shine-through measurements for the 1997 campaign at DIII-D to the expected shine-through given by theory

  9. RECENT DEVELOPMENTS ON THE 110 GHz ELECTRON CYCLOTRON INSTATLLATION ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PONCE, D.; CALLIS, R.W.; CARY, W.P.; FERRON, J.R.; GREEN, M.; GRUNLOH, H.J.; GORELOV, Y.; LOHR, J.; ELLIS, R.A.

    2002-01-01

    OAK A271 RECENT DEVELOPMENTS ON THE 110 GHZ ELECTRON CYCLOTRON INSTALLATION ON THE DIII-D TOKAMAK. Significant improvements are being implement4ed to the capability of the 110 GHz electron cyclotron system on the DIII-D tokamak. Chief among these is the addition of the fifth and sixth 1 MW class gyrotrons, increasing the power available for auxiliary heating and current drive by nearly 60%. These tubes use artificially grown diamond rf output windows to obtain high power with long pulse capability. The beams from these tubes are nearly Gaussian, facilitating coupling to the waveguide. A new fully articulating dual launcher capable of high speed spatial scanning has been designed and tested. The launcher has two axis independent steering for each waveguide. the mirrors can be rotated at up to 100 o /s. A new feedback system linking the DIII-D Plasma Control System (PCS) with the gyrotron beam voltage waveform generators permits real-time feedback control of some plasma properties such as electron temperature. The PCS can use a variety of plasma monitors to generate its control signal, including electron cyclotron emission and Mirnov probes. Electron cyclotron heating and electron cyclotron current drive (ECH and ECCD) were used during this year's DIII-D experimental campaign to control electron temperature, density, and q profiles, induce an ELM-free H-mode, and suppress the m=2/n=1 neoclassical tearing mode. The new capabilities have expanded the role of EC systems in tokamak plasma control

  10. Mechanical strength evaluation of the welded bellows for the ports of the JT-60 vacuum vessel

    International Nuclear Information System (INIS)

    Takatso, H.; Shimizu, M.; Yamamoto, M.

    1983-01-01

    Mechanical strength of the welded bellows for the ports of the JT-60 vacuum vessel was evaluated, laying the emphasis on the fatigue strength under the torsional electromagnetic force. The welded bellows were designed to be loaded with the forced deflection due to the relative displacement between the vacuum vessel and the external fixed point, the atmospheric pressure and the forced torsional angle due to the electromagnetic force. Stresses caused by the former two were estimated following the formulae proposed by the Kellogg Company. On the other hand, two formulae were established to estimate the stress caused by the last, after examining experimentally the behavior of the welded bellows under the torsional load; one is the shearing stress evaluation formula and the other is the axial bending stress evaluation formula. It was found that the welded bellows can easily buckle under the torsional load and the former formula corresponds to the case of non-buckling and the latter to the case of buckling. The present mechanical strength evaluation method was applied to the three kinds of the welded bellows to be used in the ports of the JT-60 vacuum vessel (neutral beam injection ports, vacuum pumping ports and the adjustable limiter ports) and it was confirmed that they have sufficient strength in the range of the design load conditions

  11. ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    1995-01-01

    ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers ampersand Constructors and Chicago Bridge ampersand Iron (Raytheon/CB ampersand I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB ampersand I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB ampersand I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB ampersand I and documented accordingly

  12. RECENT DEVELOPMENTS IN ALTERNATIVES TO CAMAC FOR DATA ACQUISITION AT DIII-D

    International Nuclear Information System (INIS)

    KELLMAN, D.H.; CAMPBELL, G.L.; FERRON, J.R.; PIGLOWSKI, D.A.; AUSTIN, M.E.; MCKEE, G.R.

    2004-03-01

    OAK-B135 For over twenty years, data acquisition hardware at DIII-D has been based on the CAMAC platform. These rugged and reliable systems, however, are gradually becoming obsolete due to end-of-life issues, ever-decreasing industry support of older hardware, and the availability of modern alternative hardware with superior performance. Efforts are underway at DIII-D to adopt new data acquisition solutions which exploit modern technologies and surpass the limitations of the CAMAC standard. These efforts have involved the procurement and development of data acquisition systems based on the PCI and Compact-PCI platform standards. These systems are comprised of rack-mount computers containing data acquisition boards (digitizers), Ethernet connectivity, and the drivers and software necessary for control. Each digitizer contains analog-to-digital converters, control circuitry, firmware and memory to collect, store, and transfer waveform data acquired using internal or external triggers and clocks. Software has been developed which allows DIII-D computers to program the operational parameters of the digitizers, as well as to upload acquired data into the DIII-D acquisition database. All communication between host computers and the new acquisition systems occurs via standard Ethernet connections, a vast improvement over the slower, serial loop highways used for control and data transfer with CAMAC systems. In addition, the capabilities available in modern integrated and printed circuit manufacture result in digitizers with high channel count and memory density. Cost savings are also realized by utilizing a platform based on standards of the personal computer industry. Details of the new systems at DIII-D are presented, along with initial experience with their use, and plans for future expansion and improvement

  13. Lower hybrid current drive for edge current density modification in DIII-D: Final status report

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Porkolab, M.

    1993-01-01

    Application of Lower Hybrid (LH) Current Drive (CD) in the DIII-D tokamak has been studied at LLNL, off and on, for several years. The latest effort began in February 1992 in response to a letter from ASDEX indicating that the 2.45 GHz, 3 MW system there was available to be used on another device. An initial assessment of the possible uses for such a system on DIII-D was made and documented in September 1992. Multiple meetings with GA personnel and members of the LH community nationwide have occurred since that time. The work continued through the submission of the 1995 Field Work Proposals in March 1993 and was then put on hold due to budget limitations. The purpose of this document is to record the status of the work in such a way that it could fairly easily be restarted at a future date. This document will take the form of a collection of Appendices giving both background and the latest results from the FY 1993 work, connected by brief descriptive text. Section 2 will describe the final workshop on LHCD in DIII-D held at GA in February 1993. This was an open meeting with attendees from GA, LLNL, MIT and PPPL. Summary documents from the meeting and subsequent papers describing the results will be included in Appendices. Section 3 will describe the status of work on the use of low frequency (2.45 GHZ) LH power and Parametric Decay Instabilities (PDI) for the special case of high dielectric in the edge regions of the DIII-D plasma. This was one of the critical issues identified at the workshop. Other potential issues for LHCD in the DIII-D scenarios are: (1) damping of the waves on fast ions from neutral beam injection, (2) runaway electrons in the low density edge plasma, (3) the validity of the WKB approximation used in the ray-tracing models in the steep edge density gradients

  14. Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

    2000-12-01

    Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

  15. Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, California 90095 (United States); Garofalo, A M [Columbia University, New York, New York 10027 (United States); Greenfield, C M [General Atomics, San Diego, California 92186-5608 (United States); Kaye, S M [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Menard, J E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Murakami, M [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sabbagh, S A [Columbia University, New York, New York 10027 (United States); Austin, M E [University of Texas-Austin, Austin, Texas 78712 (United States); Bell, R E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Burrell, K H [General Atomics, San Diego, California 92186-5608 (United States); Ferron, J R [General Atomics, San Diego, California 92186-5608 (United States); Gates, D A [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Groebner, R J; Hyatt, A W; Luce, T C; Petty, C C; Wade, M R; Waltz, R E [General Atomics, San Diego, California 92186-5608 (United States); Jayakumar, R J [Lawrence Livermore National Lab., Livermore, California 94550 (United States); Kinsey, J E [Lehigh Univ., Bethlehem, Pennsylvania 18015 (United States); LeBlanc, B P [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); McKee, G R [Univ. of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Okabayashi, M [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Peng, Y-K M [Oak Ridge National Lab., Oak Ridge, Tennessee 37831 (United States); Politzer, P A [General Atomics, San Diego, California 92186-5608 (United States); Rhodes, T L [Dept. of Electrical Engineering and PSTI, Univ. of California, Los Angeles, California 90095 (United States)

    2006-12-15

    Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f{sub BS} {>=} 60%, q{sub 95} {approx} 4-5 and G {identical_to}{beta}{sub N}H{sub scaling}/q{sub 95}{sup 2} {>=}0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E ? B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at {beta}{sub N} {>=} 4 with q{sub min} {>=} 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. {beta}{sub N} = 4, H{sub 89} = 2.5, with f{sub BS} > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q {>=} 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

  16. Tools for remote collaboration on the DIII-D national fusion facility

    International Nuclear Information System (INIS)

    McHarg, B.B. Jr.; Greenwood, D.

    1999-01-01

    The DIII-D national fusion facility, a tokamak experiment funded by the US Department of Energy and operated by General Atomics (GA), is an international resource for plasma physics and fusion energy science research. This facility has a long history of collaborations with scientists from a wide variety of laboratories and universities from around the world. That collaboration has mostly been conducted by travel to and participation at the DIII-D site. Many new developments in the computing and technology fields are now facilitating collaboration from remote sites, thus reducing some of the needs to travel to the experiment. Some of these developments include higher speed wide area networks, powerful workstations connected within a distributed computing environment, network based audio/video capabilities, and the use of the world wide web. As the number of collaborators increases, the need for remote tools become important options to efficiently utilize the DIII-D facility. In the last two years a joint study by GA, Princeton Plasma Physics Laboratory (PPPL), Lawrence Livermore National Laboratory (LLNL), and Oak Ridge National Laboratory (ORNL) has introduced remote collaboration tools into the DIII-D environment and studied their effectiveness. These tools have included the use of audio/video for communication from the DIII-D control room, the broadcast of meetings, use of inter-process communication software to post events to the network during a tokamak shot, the creation of a DCE (distributed computing environment) cell for creating a common collaboratory environment, distributed use of computer cycles, remote data access, and remote display of results. This study also included sociological studies of how scientists in this environment work together as well as apart. (orig.)

  17. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    Baker, D.

    1993-05-01

    The DIII-D tokamak research program is carried out by, General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data needed by International Thermonuclear Experimental Reactor (ITER) and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics

  18. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    Simonen, T.C.; Baker, D.

    1993-01-01

    The DIII-D tokamak research program is carried out by General Atomics for the U.S. Department of Energy. The DIII-D is the most flexible and best diagnosed tokamak in the world and the second largest tokamak in the U.S. The primary goal of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated in three major areas: Tokamak Physics, Divertor and Boundary Physics, and Advanced Tokamak Studies

  19. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D. [ed.

    1993-05-01

    The DIII-D tokamak research program is carried out by, General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data needed by International Thermonuclear Experimental Reactor (ITER) and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics.

  20. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    Energy Technology Data Exchange (ETDEWEB)

    Houry, M., E-mail: Michael.houry@cea.fr [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H. [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Kammerer, N.; Measson, Y. [CEA, LIST, F-92265 Fontenay-aux-Roses (France); Carrel, F.; Schoepff, V. [CEA, LIST, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  1. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    International Nuclear Information System (INIS)

    Houry, M.; Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H.; Kammerer, N.; Measson, Y.; Carrel, F.; Schoepff, V.

    2011-01-01

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  2. Design and Structural Analysis for the Vacuum Vessel of Superconducting Tokamak JT-60SC

    International Nuclear Information System (INIS)

    Kudo, Y.; Sakurai, S.; Masaki, K.; Urata, K.; Sasajima, T.; Matsukawa, M.; Sakasai, A.; Ishida, S.

    2003-01-01

    A modification of the JT-60 is planned to be a superconducting tokamak (JT-60SC) in order to establish steady-state operation of high beta plasma for 100 s, and to ensure the applicability of ferritic steel as a reduced activation material for reactor relevant break-even class plasmas. This paper describes the detailed design of the vacuum vessel, which has a unique structure for cost effective manufacturing, as well as structural analysis results for a feasibility study

  3. Studies on structural analysis related to the design of the JT-60 vacuum vessel

    International Nuclear Information System (INIS)

    Takatsu, Hideyuki

    1987-06-01

    Studies on structural analysis of a vacuum vessel of tokamak-type fusion devices are presented. The present studies are proposals for the structural analysis procedures of the tokamak-type fusion devices and are composed of five parts, each of which covers the fundamental area required for the structural analysis and design; stress analysis, dynamic response analysis, fatigue evaluation, buckling analysis and seismic analysis. Special attention is paid to the critical component, bellows and the critical load, electromagnetic forces. A new finite element method modeling technique is proposed for the stress analysis of U-shaped bellows, where the bellows is replaced by an orthotropic plate having the same stiffness as the bellows. The applicability of the present modeling technique is confirmed by verification tests. Dynamic response and fatigue of the vacuum vessel are critical issues of the structural analysis and design of the tokamak-type fusion devices. Detailed dynamic response analyses of the JT-60 vacuum vessel are presented paying special attention to the dynamic behavior of the U-shaped bellows, where the above-mentioned modeling technique of the U-shaped bellows is applied. A fatigue evaluation method of the vacuum vessel under the dynamic electromagnetic forces is proposed, which utilizes the results of the detailed dynamic response analysis. In the present method, fatigue evaluation method for random loads is applied. Torsional fatigue strength of the welded bellows is experimentally evaluated aiming the application to the port of the fusion device and it is shown that the welded bellows reveals elastic buckling and spiral distortion under a small angle of tortion. Two formulae are proposed to evaluate the stress of the welded bellows under the forced angle of tortion. (author)

  4. Stress analysis of a double-wall vacuum vessel for ITER

    International Nuclear Information System (INIS)

    Conner, D.L.; Williamson, D.E.; Nelson, B.E.

    1991-01-01

    The preliminary structural analyses performed in support of the design of the vacuum vessel for the International Thermonuclear Experimental Reactor (ITER) are described. A thin, double-wall, all-welded structure is the proposed design concept analyzed. The results of the static stress analysis indicate the adequacy of such a structure. The effects of the proposed high-aspect-ratio design configuration on loading and stresses are also discussed. 4 refs., 6 figs., 1 tab

  5. Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    Iida, H.; Minato, A.; Kitamura, K.

    1983-01-01

    The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li 2 O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

  6. Dynamic response of the JT-60 vacuum vessel under the electromagnetic forces

    International Nuclear Information System (INIS)

    Takatsu, H.; Shimizu, M.; Ohta, M.

    1982-01-01

    Dynamic response analyses of the JAERI Tokamak 60 (JT-60) vacuum vessel were carried out under three kinds of saddle-like electromagnetic forces. In the analysis, the dynamic response of the bellows was obtained by dividing it into three components; the first, caused by the forced deflection due to the displacement of an adjacent rigid ring; the second, caused by inertia force; and the third, caused by a saddle-like electromagnetic force. Eigenvalue analyses showed that the 20th mode is a typical rotation mode of the rigid ring around the major radius with a natural frequency of 46.3 Hz. From the results of the dynamic response analyses, the maximum displacement response of the rigid ring was 3.1 mm and remarkable dynamic response was observed in the case of plasma disruption with a time constant of 1 ms. In cases of start-up of the plasma current and plasma disruption with a time constant of 50 ms, the rigid ring vibrates quasi-statically. It is clear that the dynamic behavior of the vacuum vessel is governed mainly by the saddle-like electromagnetic force, with a smaller effect of the inverse saddle-like electromagnetic force on the dynamic response of the vacuum vessel. (orig.)

  7. Experimental and analytical investigations to air and steam ingress into the vacuum vessel of fusion reactors

    International Nuclear Information System (INIS)

    Kruessenberg, A.K.

    1996-12-01

    The basic fusion safety objective is the development of fusion power plants with features that protect individuals, society and the environment by establishing and maintaining an effective defence against radiological and other hazards. The most important specific principle is the establishment of three sequential levels of defence, characterized in priority order by prevention, protection and mitigation. The safety conscious selection of materials as one prevention feature gives the basis for the work described in this report. In order to protect the metallic first wall of fusion reactors from direct interaction with the plasma an extra armour is foreseen. Carbon offers the features low atomic number, high melting point, high thermal conductivity and good mechanical stability up to high temperatures making it to a favourite armour material. Looking on the safety behaviour of fusion reactors it has to be noted that carbon is unstable against oxidizing media like oxygen and steam at high temperatures und carbon has a high sorption capacity for radiologically important tritium. And tritium used as intermediate fuel in the actual reactor concepts is the one form radioactivity is present in fusion reactors. Accidents like loss of vacuum (LOVA) will lead to an air ingress into the vacuum vessel, oxidation of the hot carbon and a partial mobilization of the sorbed tritium. In a similar manner loss of coolant into vacuum (LOCIV) will lead to a water/steam ingress into the vacuum vessel, also accompanied by carbon oxidation and tritium release. (orig.)

  8. Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, C. M., E-mail: coopercm@fusion.gat.com [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Pace, D. C.; Paz-Soldan, C.; Eidietis, N. W. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Commaux, N.; Shiraki, D. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California, San Diego, La Jolla, California 92093-0533 (United States)

    2016-11-15

    A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5–100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead “pinhole camera” mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.

  9. Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

    International Nuclear Information System (INIS)

    Cooper, C. M.; Pace, D. C.; Paz-Soldan, C.; Eidietis, N. W.; Commaux, N.; Shiraki, D.; Hollmann, E. M.

    2016-01-01

    A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5–100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead “pinhole camera” mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.

  10. COMPLETE SUPPRESSION OF THE m=2/n-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

    International Nuclear Information System (INIS)

    PETTY, CC; LAHAYE, LA; LUCE, TC; HUMPHREYS, DA; HYATT, AW; PRATER, R; STRAIT, EJ; WADE, MR

    2003-01-01

    A271 COMPLETE SUPPRESSION OF THE M=2/N-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. The first suppression of the important and deleterious m=2/n=1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace the ''missing'' bootstrap current in the island O-point. Experiments on the DIII-D tokamak verify the maximum shrinkage of the m=2/n=1 island occurs when the ECCD location coincides with the q = 2 surface. The DIII-D plasma control system is put into search and suppress mode to make small changes in the toroidal field to find and lock onto the optimum position, based on real time measurements of dB θ /dt, for complete m=2/n=1 NTM suppression by ECCD. The requirements on the ECCD for complete island suppression are well modeled by the modified Rutherford equation for the DIII-D plasma conditions

  11. Temperature distributions in a Tokamak vacuum vessel of fusion reactor after the loss-of-vacuum-events occurred

    International Nuclear Information System (INIS)

    Takase, K.; Kunugi, T.; Shibata, M.; Seki, Y.

    1998-01-01

    If a loss-of-vacuum-event (LOVA) occurred in a fusion reactor, buoyancy-driven exchange flows would occur at breaches of a vacuum vessel (VV) due to the temperature difference between the inside and outside of the VV. The exchange flows may bring mixtures of activated materials and tritium in the VV to the outside through the breaches, and remove decay heat from the plasma-facing components of the VV. Therefore, the LOVA experiments were carried out under the condition that one or two breaches was opened and that the VV was heated to a maximum 200 C, using a small-scaled LOVA experimental apparatus. Air and helium gas were provided as working fluids. Fluid and wall temperature distributions in the VV were measured and the flow patterns in the VV were estimated by using these temperature distributions. It was found that: (1) the exchange mass in the VV depended on the breach positions; (2) the exchange flow at the single breach case became a counter-current flow when the breach was at the roof of the VV and a stratified flow when it was at the side wall; (3) and that at the double breach case, a one-way flow between two breaches was formed. (orig.)

  12. Suppression of large edge localized modes with a stochastic magnetic boundary in high confinement DIII-D plasmas

    International Nuclear Information System (INIS)

    Evans, T.E.; Moyer, R.A.; Watkins, J.G.

    2005-01-01

    Large sub-millisecond heat pulses due to Type-I ELMs have been eliminated reproducibly in DIII.D for periods approaching 7 energy confinement times with small dc currents driven in a simple magnetic perturbation coil. The current required to eliminate all but a few isolated Type-I ELM impulses during a coil pulse is less than 0.4% of plasma current. Based on vacuum magnetic field line modeling, the perturbation fields resonate strongly with plasma flux surfaces across most of the pedestal region (0.9 ≤ Ψ N ≤ 1.0) when q 95 = 3.7±0.2 creating small remnant magnetic islands surrounded by weakly stochastic field lines. The stored energy, β N , H-mode quality factor and global energy confinement time are unaltered. Although some isolated ELM-like events typically occur, long periods free of large Type-I ELMs (Δt > 4-6 τ E ) have been reproduced numerous times, on multiple experimental run days including cases matching the ITER scenario 2 flux surface shape. Since large Type-I ELM impulses represent a severe constraint on the survivability of the divertor target plates in future fusion devices such as ITER, a proven method of eliminating these impulses is critical for the development of tokamak reactors. Results presented in this paper indicate that non-axisymmetric edge magnetic perturbations could be a promising option for controlling ELMs in future tokamaks such as ITER. (author)

  13. Initial results of high resolution L-H transition studies on DIII-D

    International Nuclear Information System (INIS)

    Wang, G; Rhodes, T L; Doyle, E J; Peebles, W A; Zeng, L; Burrell, K H; McKee, G R; Groebner, R J; Evans, T E

    2004-01-01

    Understanding the L-H transition in tokamaks has been an important area of research for more than two decades. High time resolution diagnostics on DIII-D allow detailed characterization of the L-H transition and, therefore, testing and benchmarking of theoretical models. An experiment was performed in DIII-D utilizing a novel, high temporal and spatial resolution reflectometer density profile system to measure densities from the SOL to the inside separatrix. Initial data analysis indicates different density profile evolution during L-H transitions in upper single-null and lower single-null divertor configuration plasmas. A detailed comparison of the density gradient and fluctuation changes is presented for these two cases

  14. Lower-hybrid counter current drive for edge current density modification in DIII-D

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Nevins, W.M.; Porkolab, M.; Bonoli, P.T.; Harvey, R.W.

    1994-01-01

    Each of the Advanced Tokamak operating modes in DIII-D is thought to have a distinctive current density profile. So far these modes have only been achieved transiently through experiments which ramp the plasma current and shape. Extension of these modes to steady state requires non-inductive current profile control, e.g., with lower hybrid current drive (LHCD). Calculations of LHCD have been done for DIII-D using the ACCOME and CQL3D codes, showing that counter driven current at the plasma edge can cancel some of the undesirable edge bootstrap current and potentially extend the VH-mode. Results will be presented for scenarios using 2.45 GHz LH waves launched from both the midplane and off-axis ports. The sensitivity of the results to injected power, n e and T e , and launched wave spectrum will also be shown

  15. Edge density fluctuation diagnostic for DIII-D using lithium beams: 1992 annual report

    International Nuclear Information System (INIS)

    Thomas, D.M.

    1994-01-01

    During the past several months the Lithium beam diagnostic was commissioned of DIII-D and began yielding useful information. The author developed the remote control and monitoring of the ion source operation and beam formation and focussing, and integrated the control system and data acquisition into the DIII-D operating system. Several detector types were fabricated, and fluorescence data were collected using several differing detector arrangements. Beam-gas measurements were conducted to analyze the intrinsic beam fluctuations and stability. Fluorescence data was then obtained on a number of Tokamak discharges under varying discharge conditions. Analysis of this initial data is proceeding but has already yielded some interesting features. These include changes in the edge plasma density behavior during the l- to h-transition, disruptions, and edge localized modes (ELMs). Based on the quality of data obtained the author proceeded with the design and construction of the full 16-channel detection system which will be completed and tested shortly

  16. Reduction of recycling in DIII-D by degassing and conditioning of the graphite tiles

    International Nuclear Information System (INIS)

    Jackson, G.L.; Taylor, T.S.; Allen, S.L.

    1988-05-01

    Reduced recycling, reduced edge neutral pressure, improved density control, and improved discharge reproducibility have been achieved in the DIII-D tokamak by in situ helium conditioning of the graphite tiles. An improvement in energy confinement has been observed in hydrogen discharges with hydrogen beam injection after helium preconditioning. After the graphite wall coverage in DIII-D was increased to 40%, helium glow wall conditioning, routinely applied before each tokamak discharge, has been necessary to reduce recycling and obtain H-mode. The utilization of helium glow wall conditioning was an important factor in the achievement of an ohmic H-mode, i.e. no auxillary heating, with significant improvement in ohmic energy confinement. 16 refs., 8 figs

  17. Feedback Control of Resistive Wall Modes in Slowly Rotating DIII-D Plasmas

    Science.gov (United States)

    Okabayashi, M.; Chance, M. S.; Takahashi, H.; Garofalo, A. M.; Reimerdes, H.; in, Y.; Chu, M. S.; Jackson, G. L.; La Haye, R. J.; Strait, E. J.

    2006-10-01

    In slowly rotating plasmas on DIII-D, the requirement of RWM control feedback have been identified, using a MHD code along with measured power supply characteristics. It was found that a small time delay is essential for achieving high beta if no rotation stabilization exists. The overall system delay or the band pass time constant should be in the range of 0.4 of the RWM growth time. Recently the control system was upgraded using twelve linear audio amplifiers and a faster digital control system, reducing the time-delay from 600 to 100 μs. The advantage has been clearly observed when the RWMs excited by ELMs were effectively controlled by feedback even if the rotation transiently slowed nearly to zero. This study provides insight on stability in the low- rotation plasmasw with balanced NBI in DIII-D and also in ITER.

  18. Handling and archiving of magnetic fusion data at DIII-D

    International Nuclear Information System (INIS)

    VanderLaan, J.F.; Miller, S.; McHarg, B.B. Jr.; Henline, P.A.

    1995-10-01

    Recent modifications to the computer network at DIII-D enhance the collection and distribution of newly acquired and archived experimental data. Linked clients and servers route new data from diagnostic computers to centralized mass storage and distribute data on demand to local and remote workstations and computers. Capacity for data handling exceeds the upper limit of DIII-D Tokamak data production of about 4 GBytes per day. Network users have fast access to new data stored on line. An interactive program handles requests for restoration of data archived off line. Disk management procedures retain selected data on line in preference to other data. Redundancy of all components on the archiving path from the network to magnetic media has prevented loss of data. Older data are rearchived as dictated by limited media life

  19. PROGRESS TOWARD FULLY NONINDUCTIVE, HIGH BETA DISCHARGES IN DIII-D

    International Nuclear Information System (INIS)

    GREENFIELD, CM; FERRON, JR; MURAKAMI, M; WADE, MR; BUDNY, RV; BURRELL, KH; CASPER, TA; DeBOO, JC; DOYLE, EJ; GAROFALO, AM; JAYAKUMAR, RJ; KESSEL, C; LAO, LL; LOHR, J; LUCE, TC; MAKOWSKI, MA; MENARD, JE; PETRIE, TW; PETTY, CC; PINSKER, RI; PRATER, R; POLITZER, PA; St JOHN, HE; TAYLOR, TS; WEST, WP; DIII-D NATIONAL TEAM

    2003-01-01

    OAK-B135 Advanced Tokamak (AT) research in DIII-D focuses on developing a scientific basis for steady-state, high performance operation. For optimal performance, these experiments routinely operate with β above the n = 1 no-wall limit, enabled by active feed-back control. The ideal wall β limit is optimized by modifying the plasma shape, current and pressure profile. Present DIII-D AT experiments operate with f BS ∼ 50%-60%, with a long-term goal of ∼ 90%. Additional current is provided by neutral beam and electron cyclotron current drive, the latter being localized well away from the magnetic axis (ρ ∼ 0.4-0.5). Guided by integrated modeling, recent experiments have produced discharges with β ∼ 3%, β N ∼ 3, f BS ∼ 55% and noninductive fraction f NI ∼ 90%. Additional control is anticipated using fast wave current drive to control the central current density

  20. The production and confinement of runaway electrons with impurity ''killer'' pellets in DIII-D

    International Nuclear Information System (INIS)

    Evans, T.E.; Taylor, P.L.; Whyte, D.G.

    1998-12-01

    Prompt runaway electron bursts, generated by rapidly cooling DIII-D plasmas with argon killer pellets, are used to test a recent knock-on avalanche theory describing the growth of multi-MeV runaway electron currents during disruptions in tokamaks. Runaway current amplitudes, observed during some but not all DIII-D current quenches, are consistent with growth rates predicted by the theory assuming a pre-current quench runaway electron density of approximately 10 15 m -3 . Argon killer pellet modeling yields runaway densities of between 10 15 --10 16 m -3 in these discharges. Although knock-on avalanching appears to agree rather well with the measurements, relatively small avalanche amplification factors combined with uncertainties in the spatial distribution of pellet mass and cooling rates make it difficult to unambiguously confirm the proposed theory with existing data

  1. Scintillator-based diagnostic for fast ion loss measurements on DIII-D

    International Nuclear Information System (INIS)

    Fisher, R. K.; Van Zeeland, M. A.; Pace, D. C.; Heidbrink, W. W.; Muscatello, C. M.; Zhu, Y. B.; Garcia-Munoz, M.

    2010-01-01

    A new scintillator-based fast ion loss detector has been installed on DIII-D with the time response (>100 kHz) needed to study energetic ion losses induced by Alfven eigenmodes and other MHD instabilities. Based on the design used on ASDEX Upgrade, the diagnostic measures the pitch angle and gyroradius of ion losses based on the position of the ions striking the two-dimensional scintillator. For fast time response measurements, a beam splitter and fiberoptics couple a portion of the scintillator light to a photomultiplier. Reverse orbit following techniques trace the lost ions to their possible origin within the plasma. Initial DIII-D results showing prompt losses and energetic ion loss due to MHD instabilities are discussed.

  2. Dimensionally similar discharges with central rf heating on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Luce, T.C.; Pinsker, R.I.

    1993-04-01

    The scaling of L-mode heat transport with normalized gyroradius is investigated on the DIII-D tokamak using central rf heating. A toroidal field scan of dimensionally similar discharges with central ECH and/or fast wave heating show gyro-Bohm-like scaling both globally and locally. The main difference between these restats and those using NBI heating on DIII-D is that with rf heating the deposition profile is not very sensitive to the plasma density. Therefore central heating can be utilized for both the low-B and high-B discharges, whereas for NBI the power deposition is decidedly off-axis for the high-B discharge (i.e., high density)

  3. Exposures of tungsten nanostructures to divertor plasmas in DIII-D

    International Nuclear Information System (INIS)

    Rudakov, D L; Doerner, R P; Baldwin, M J; Boedo, J A; Hollmann, E M; Moyer, R A; Wong, C P C; Chrobak, C P; Guo, H Y; Leonard, A W; Pace, D C; Thomas, D M; Wright, G M; Abrams, T; Briesemeister, A R; McLean, A G; Fenstermacher, M E; Lasnier, C J; Watkins, J G

    2016-01-01

    Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with little obvious damage except in the areas where unipolar arcing occurred. Arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces. (paper)

  4. Initial results of high resolution L-H transition studies on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Wang, G [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Rhodes, T L [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Peebles, W A [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Zeng, L [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186 (United States); McKee, G R [University of Wisconsin-Madison, 1500 Engineering Drive, Madison, WI 53706 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186 (United States); Evans, T E [General Atomics, PO Box 85608, San Diego, CA 92186 (United States)

    2004-05-01

    Understanding the L-H transition in tokamaks has been an important area of research for more than two decades. High time resolution diagnostics on DIII-D allow detailed characterization of the L-H transition and, therefore, testing and benchmarking of theoretical models. An experiment was performed in DIII-D utilizing a novel, high temporal and spatial resolution reflectometer density profile system to measure densities from the SOL to the inside separatrix. Initial data analysis indicates different density profile evolution during L-H transitions in upper single-null and lower single-null divertor configuration plasmas. A detailed comparison of the density gradient and fluctuation changes is presented for these two cases.

  5. Operational upgrades to the DIII-D 60 GHz electron cyclotron resonant heating system

    International Nuclear Information System (INIS)

    Harris, T.E.; Cary, W.P.

    1993-10-01

    One of the primary components of the DIII-D radio frequency (rf) program over the past seven years has been the 60 GHz electron cyclotron resonant heating (ECRH) system. The system now consists of eight units capable of operating and controlling eight Varian VGE-8006 60 GHz, 200 kW gyrotrons along with their associated waveguide components. This paper will discuss the operational upgrades and the overall system performance. Many modifications were instituted to enhance the system operation and performance. Modifications discussed in this paper include an improved gyrotron tube-fault response network, a computer controlled pulse-timing and sequencing system, and an improved high-voltage power supply control interface. The discussion on overall system performance will include operating techniques used to improve system operations and reliability. The techniques discussed apply to system start-up procedures, operating the system in a conditioning mode, and operating the system during DIII-D plasma operations

  6. Effects of particle exhaust on neutral compression ratios in DIII-D

    International Nuclear Information System (INIS)

    Colchin, R.J.; Maingi, R.; Wade, M.R.; Allen, S.L.; Greenfield, C.M.

    1998-08-01

    In this paper, neutral particles in DIII-D are studied via their compression in the plenum and via particle exhaust. The compression of gas in the plena is examined in terms of the magnetic field configuration and wall conditions. DIII-D compression ratios are observed in the range from 1 to ≥ 1,000. Particle control ultimately depends on the exhaust of neutrals via plenum or wall pumping. Wall pumping or outgassing is calculated by means of a detailed particle balance throughout individual discharges, and its effect on particle control is discussed. It is demonstrated that particle control through wall conditioning leads to lower normalized densities. A two-region model shows that the gas compression ratio (C div = divertor plenum neutral pressure/torus neutral pressure) can be interpreted in relation to gas flows in the torus and divertor including the pumping speed of the plenum cryopumps, plasma pumping, and the pumping or outgassing of the walls

  7. Faraday-effect polarimeter diagnostic for internal magnetic field fluctuation measurements in DIII-D.

    Science.gov (United States)

    Chen, J; Ding, W X; Brower, D L; Finkenthal, D; Muscatello, C; Taussig, D; Boivin, R

    2016-11-01

    Motivated by the need to measure fast equilibrium temporal dynamics, non-axisymmetric structures, and core magnetic fluctuations (coherent and broadband), a three-chord Faraday-effect polarimeter-interferometer system with fast time response and high phase resolution has recently been installed on the DIII-D tokamak. A novel detection scheme utilizing two probe beams and two detectors for each chord results in reduced phase noise and increased time response [δb ∼ 1G with up to 3 MHz bandwidth]. First measurement results were obtained during the recent DIII-D experimental campaign. Simultaneous Faraday and density measurements have been successfully demonstrated and high-frequency, up to 100 kHz, Faraday-effect perturbations have been observed. Preliminary comparisons with EFIT are used to validate diagnostic performance. Principle of the diagnostic and first experimental results is presented.

  8. Faraday-effect polarimeter diagnostic for internal magnetic field fluctuation measurements in DIII-D

    International Nuclear Information System (INIS)

    Chen, J.; Ding, W. X.; Brower, D. L.; Finkenthal, D.; Muscatello, C.; Taussig, D.; Boivin, R.

    2016-01-01

    Motivated by the need to measure fast equilibrium temporal dynamics, non-axisymmetric structures, and core magnetic fluctuations (coherent and broadband), a three-chord Faraday-effect polarimeter-interferometer system with fast time response and high phase resolution has recently been installed on the DIII-D tokamak. A novel detection scheme utilizing two probe beams and two detectors for each chord results in reduced phase noise and increased time response [δb ∼ 1G with up to 3 MHz bandwidth]. First measurement results were obtained during the recent DIII-D experimental campaign. Simultaneous Faraday and density measurements have been successfully demonstrated and high-frequency, up to 100 kHz, Faraday-effect perturbations have been observed. Preliminary comparisons with EFIT are used to validate diagnostic performance. Principle of the diagnostic and first experimental results is presented.

  9. A decade of DIII-D research. Final report for the period of work, October 1, 1989--September 30, 1998

    International Nuclear Information System (INIS)

    1999-03-01

    During the ten-year DIII-D tokamak operating period of 1989 through 1998, major scientific advances and discoveries were made and facility upgrades and improvements were implemented. Each year, annual reports as well as journal and international conference proceedings document the year-by-year advances (summarized in Section 7). This final contract report, provides a summary of these historical accomplishments. Section 2 encapsulates the 1998 status of DIII-D Fusion Science research. Section 3 summarizes the DIII-D facility operations. Section 4 describes the major upgrades to the DIII-D facility during this period. During the ten-year period, DIII-D has grown from predominantly a General Atomics program to a national center for fusion science with participants from over 50 collaborating institutions and 300 users who spend more than one week annually at DIII-D to carry out experiments or data analysis. In varying degrees, these collaborators participate in formulating the research program directions. The major collaborating institution programs are described in Section 6

  10. A decade of DIII-D research. Final report for the period of work, October 1, 1989--September 30, 1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    During the ten-year DIII-D tokamak operating period of 1989 through 1998, major scientific advances and discoveries were made and facility upgrades and improvements were implemented. Each year, annual reports as well as journal and international conference proceedings document the year-by-year advances (summarized in Section 7). This final contract report, provides a summary of these historical accomplishments. Section 2 encapsulates the 1998 status of DIII-D Fusion Science research. Section 3 summarizes the DIII-D facility operations. Section 4 describes the major upgrades to the DIII-D facility during this period. During the ten-year period, DIII-D has grown from predominantly a General Atomics program to a national center for fusion science with participants from over 50 collaborating institutions and 300 users who spend more than one week annually at DIII-D to carry out experiments or data analysis. In varying degrees, these collaborators participate in formulating the research program directions. The major collaborating institution programs are described in Section 6.

  11. Non-inductive current drive experiments on DIII-D, and future plans

    International Nuclear Information System (INIS)

    Prater, R.; Austin, M.; Baity, F.W.; Callis, R.W.; Chiu, S.C.; DeGrassie, J.S.; Freeman, R.L.; Forest, C.B.; Goulding, R.H.; Harvey, R.W.; Hoffman, D.J.; Ikezi, H.; Lohr, J.; James, R.A.; Kupfer, K.; Lin-Liu, Y.R.; Luce, T.C.; Moeller, C.P.; Petty, C.C.; Pinsker, R.I.; Porkolab, M.; Squire, J.; Trukhin, V.

    1995-01-01

    Experiments on DIII-D (and other tokamaks) have shown that improved performance can follow from optimization of the current density profile. Increased confinement of energy and a higher limit on β have both been found in discharges in which the current density profile is modified through transient means, such as ramping of current or elongation. Peaking of the current distribution to obtain discharges with high internal inductance l i has been found to be beneficial. Alternatively, discharges with broader profiles, as in the VH mode or with high β poloidal, have shown improved performance. Non-inductive current drive is a means to access these modes of improved confinement on a steady state basis. Accordingly, experiments on non-inductive current drive are underway on the DIII-D tokamak using fast waves and electron cyclotron waves. Recent experiments on fast wave current drive have demonstrated the ability to drive up to 180kA of non-inductive current using 1.5MW of power at 60MHz, including the contribution from 1MW of ECCD and the bootstrap current. Higher power r.f. current drive systems are needed to affect strongly the current profile on DIII-D. An upgrade to the fast wave current drive system is underway to increase the total power to 6MW, using two additional antennas and two new 30-120MHz transmitters. Additionally, a 1MW prototype ECH system at 110GHz is being developed (with eventual upgrade to 10MW). With these systems, non-inductive current drive at the 1MA level will be available for experiments on profile control in DIII-D. ((orig.))

  12. Initial assessment of lower hybrid current drive in DIII-D

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Porkolab, M.

    1992-01-01

    In this report we discuss the choice of rf frequency for LHCD in DIII-D, and its consequences for the operating regimes. We also discuss some Brambilla code results for particular launcher geometry, as well as edge density requirements for efficient coupling by the launcher. We present results of the ACCOME code modelling for selected frequency spectra. Finally, a summary is given, as well as suggestions for future studies

  13. 2 MW 110 GHz ECH heating system for DIII-D

    International Nuclear Information System (INIS)

    Moeller, C.; Prater, R.; Callis, R.; Remsen, D.; Doane, J.; Cary, W.; Phelps, R.; Tupper, M.

    1990-09-01

    A 2 MW 110 GHz ECH system using Varian 0.5 MW gyrotrons is under construction for use on the DIII-D tokamak by late 1991. Most of the components are being design and fabricated at General Atomics, including the gyrotron tanks, superconducting magnets, and transmission line. These components are intended for operation with 10 second pulses and, in the future, with 1 MW gyrotrons. 6 refs., 5 figs

  14. DIII-D Integrated plasma control solutions for ITER and next-generation tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Ferron, J.R.; Hyatt, A.W.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; In, Y.

    2008-01-01

    Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed 'integrated plasma control,' enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple ('control-level') models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks

  15. DIII-D RESEARCH OPERATIONS ANNUAL REPORT TO THE U.S. DEPARTMENT OF ENERGY

    Energy Technology Data Exchange (ETDEWEB)

    EVANS,TE

    2003-12-01

    OAK-B135 The mission of the DIII-D research program is: ''To establish the scientific basis for the optimization of the tokamak approach to fusion energy production. The program is focused on developing the ultimate potential of the tokamak by building a better fundamental understanding of the physics of plasma confinement, stability, current drive and heating in high performance discharges while utilizing new scientific discoveries and improvements in their knowledge of these basic areas to create more efficient control systems, improved plasma diagnostics and to identify new types of enhanced operating regimes with improved stability properties. In recent years, this development path has culminated in the advanced tokamak (AT) approach. An approach that has shown substantial promise for improving both the fusion yield and the energy density of a burning plasma device. While the challenges of increasing AT plasma performance levels with greater stability for longer durations are significant, the DIII-D program has an established plan that brings together both the critical resources and the expertise needed to meet these challenges. The DIII-D research staff is comprised of about 300 individuals representing 60 institutions with many years of integrated research experience in tokamak physics, engineering and technology. The DIII-D tokamak is one of the most productive, flexible and best diagnosed magnetic fusion research devices in the world. It has significantly more flexibility than most tokamaks and continues to pioneer the development of sophisticated new plasma feedback control tools that enable the explorations of new frontiers in fusion science and engineering.

  16. DIII-D RESEARCH OPERATIONS ANNUAL REPORT October 1, 2001 through September 30, 2002

    International Nuclear Information System (INIS)

    EVANS, T.E.

    2003-01-01

    OAK-B135 The mission of the DIII-D research program is: ''To establish the scientific basis for the optimization of the tokamak approach to fusion energy production. The program is focused on developing the ultimate potential of the tokamak by building a better fundamental understanding of the physics of plasma confinement, stability, current drive and heating in high performance discharges while utilizing new scientific discoveries and improvements in their knowledge of these basic areas to create more efficient control systems, improved plasma diagnostics and to identify new types of enhanced operating regimes with improved stability properties. In recent years, this development path has culminated in the advanced tokamak (AT) approach. An approach that has shown substantial promise for improving both the fusion yield and the energy density of a burning plasma device. While the challenges of increasing AT plasma performance levels with greater stability for longer durations are significant, the DIII-D program has an established plan that brings together both the critical resources and the expertise needed to meet these challenges. The DIII-D research staff is comprised of about 300 individuals representing 60 institutions with many years of integrated research experience in tokamak physics, engineering and technology. The DIII-D tokamak is one of the most productive, flexible and best diagnosed magnetic fusion research devices in the world. It has significantly more flexibility than most tokamaks and continues to pioneer the development of sophisticated new plasma feedback control tools that enable the explorations of new frontiers in fusion science and engineering

  17. Stability and control of resistive wall modes in high beta, low rotation DIII-D plasmas

    International Nuclear Information System (INIS)

    Garofalo, A.M.; Jackson, G.L.; Haye, R.J. La; Okabayashi, M.; Reimerdes, H.; Strait, E.J.; Ferron, J.R.; Groebner, R.J.; In, Y.; Lanctot, M.J.; Matsunaga, G.; Navratil, G.A.; Solomon, W.M.; Takahashi, H.; Takechi, M.; Turnbull, A.D.

    2007-01-01

    Recent high-β DIII-D (Luxon J.L. 2002 Nucl. Fusion 42 64) experiments with the new capability of balanced neutral beam injection show that the resistive wall mode (RWM) remains stable when the plasma rotation is lowered to a fraction of a per cent of the Alfven frequency by reducing the injection of angular momentum in discharges with minimized magnetic field errors. Previous DIII-D experiments yielded a high plasma rotation threshold (of order a few per cent of the Alfven frequency) for RWM stabilization when resonant magnetic braking was applied to lower the plasma rotation. We propose that the previously observed rotation threshold can be explained as the entrance into a forbidden band of rotation that results from torque balance including the resonant field amplification by the stable RWM. Resonant braking can also occur naturally in a plasma subject to magnetic instabilities with a zero frequency component, such as edge localized modes. In DIII-D, robust RWM stabilization can be achieved using simultaneous feedback control of the two sets of non-axisymmetric coils. Slow feedback control of the external coils is used for dynamic error field correction; fast feedback control of the internal non-axisymmetric coils provides RWM stabilization during transient periods of low rotation. This method of active control of the n = 1 RWM has opened access to new regimes of high performance in DIII-D. Very high plasma pressure combined with elevated q min for high bootstrap current fraction, and internal transport barriers for high energy confinement, are sustained for almost 2 s, or 10 energy confinement times, suggesting a possible path to high fusion performance, steady-state tokamak scenarios

  18. Dependence of {beta} {center_dot} {tau} on plasma shape in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Lazarus, E.A. [Oak Ridge National Lab., TN (United States)

    1993-12-31

    In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product {beta} {center_dot} {tau} has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak. (author) 5 refs., 5 figs.

  19. Dependence of β·τ on plasma shape in DIII-D

    International Nuclear Information System (INIS)

    Lazarus, E.A.

    1993-05-01

    In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product β·τ has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak

  20. Characterization and Modification of Edge-Driven Instabilities in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Ferron, J.R.; Lao, L.L.; Osborne, T.H.; Strait, E.J.; Turnbull, A.D.; Miller, R.L.; Taylor, T.S.; Doyle, E.J.; Rice, B.W.; Zhang, C.; Chen, L.; Baylor, L.R.; Murakami, M.; Wade, M.R.

    1999-01-01

    The character of edge localized modes (ELMs) and the height of the edge pressure pedestal in DIII-D tokamak H-mode discharges have been modified by varying the discharge shape (triangularity and squareness) and the safety factor, increasing the edge radiation, and injecting deuterium pellets. Changes in the ELM frequency and amplitude, and the magnitude of the edge pressure gradient, and changes in the calculated extent of the region of access to the ballooning mode second stability regime are observed

  1. System control and data acquisition of the two new FWCD RF systems at DIII-D

    International Nuclear Information System (INIS)

    Harris, T.E.; Allen, J.C.; Cary, W.P. Petty, C.C.

    1995-10-01

    The Fast Wave Current Drive (FWCD) system at DIII-D has increased its available radio frequency (RF) power capabilities with the addition of two new high power transmitters along with their associated transmission line systems. A Sun Sparc-10 workstation, functioning as the FWCD operator console, is being used to control transmitter operating parameters and transmission line tuning parameters, along with acquiring data and making data available for integration into the DIII-D data acquisition system. Labview, a graphical user interface application, is used to manage and control the above processes. This paper will discuss the three primary branches of the FWCD computer control system: transmitter control, transmission line tuning control, and FWCD data acquisition. The main control program developed uses VXI, GPIB, CAMAC, Serial, and Ethernet protocols to blend the three branches together into one cohesive system. The control of the transmitters utilizes VXI technology to communicate with the transmitter's digital interface. A GPIB network allows for communication with various instruments and CAMAC crate controllers. CAMAC crates are located at each phase-shifter/stub-tuner station and are used to digitize transmission line parameters along with transmission line fault detection during RF transmission. The phase-shifter/stub-tuner stations are located through out the DIII-D facility and are controlled from the FWCD operator console via the workstation's Serial port. The Sun workstation has an Ethernet connection allowing for the utilization of the DIII-D data acquisition open-quotes Open Systemclose quotes architecture and of course providing communication with the rest of the world

  2. Simulation of enhanced tokamak performance on DIII-D using fast wave current drive

    International Nuclear Information System (INIS)

    Grassie, J.S. de; Lin-Liu, Y.R.; Petty, C.C.; Pinsker, R.I.; Chan, V.S.; Prater, R.; John, H. St.; Baity, F.W.; Goulding, R.H.; Hoffman, D.H.

    1993-01-01

    The fast magnetosonic wave is now recognized to be a leading candidate for noninductive current drive for the tokamak reactor due to the ability of the wave to penetrate to the hot dense core region. Fast wave current drive (FWCD) experiments on DIII-D have realized up to 120 kA of rf current drive, with up to 40% of the plasma current driven noninductively. The success of these experiments at 60 MHz with a 2 MW transmitter source capability has led to a major upgrade of the FWCD system. Two additional transmitters, 30 to 120 MHz, with a 2 MW source capability each, will be added together with two new four-strap antennas in early 1994. Another major thrust of the DIII-D program is to develop advanced tokamak modes of operation, simultaneously demonstrating improvements in confinement and stability in quasi-steady-state operation. In some of the initial advanced tokamak experiments on DIII-D with neutral beam heated (NBI) discharges it has been demonstrated that energy confinement time can be improved by rapidly elongating the plasma to force the current density profile to be more centrally peaked. However, this high-l i phase of the discharge with the commensurate improvement in confinement is transient as the current density profile relaxes. By applying FWCD to the core of such a κ-ramped discharge it may be possible to sustain the high internal inductance and elevated confinement. Using computational tools validated on the initial DIII-D FWCD experiments we find that such a high-l i advanced tokamak discharge should be capable of sustainment at the 1 MA level with the upgraded capability of the FWCD system. (author) 16 refs., 3 figs., 1 tab

  3. First results on fast wave current drive in advanced tokamak discharges in DIII-D

    International Nuclear Information System (INIS)

    Prater, R.; Cary, W.P.; Baity, F.W.

    1995-07-01

    Initial experiments have been performed on the DIII-D tokamak on coupling, direct electron heating, and current drive by fast waves in advanced tokamak discharges. These experiments showed efficient central heating and current drive in agreement with theory in magnitude and profile. Extrapolating these results to temperature characteristic of a power plant (25 keV) gives current drive efficiency of about 0.3 MA/m 2

  4. A neural network for the analysis of DIII-D charge exchange recombination data

    International Nuclear Information System (INIS)

    Baker, D.R.; Groebner, R.J.; Burrell, K.H.

    1994-01-01

    A neural network of the multiple-layer perceptron (MLP) type, named CERNEUR, has been created for the task of analysing the charge exchange recombination data from DIII-D for the purpose of providing control-room ion temperatures and rotation velocities between shots and, in the future, to provide initial guesses for the standard curve-fitting code. CERNEUR provides very useful 'control-room' in-between shot analysis of the rotation velocity and ion temperature profiles. (author)

  5. Recent improvements to the DIII-D neutral beam instrumentation and control system

    International Nuclear Information System (INIS)

    Kellman, D.H.; Hong, R.

    1997-11-01

    The DIII-D neutral beam (NB) instrumentation and control (I and C) system provides for operational control and synchronization of the eight DIII-D neutral beam injection systems, as well as for pertinent data acquisition and safety interlocking. Recently, improvements were made to the I and C system. With the replacement of the NB control computers, new signal interfacing was required to accommodate the elimination of physical operator panels, in favor of graphical user interface control pages on computer terminal screens. The program in the mode control (MC) programmable logic controller (PLC), which serves as a logic-processing interface between the NB control computers and system hardware, was modified to improve the availability of NB heating of DIII-D plasmas in the event that one or more individual beam systems suddenly become unavailable while preparing for a tokamak experimental shot sequences. An upgraded computer platform was adopted for the NB control system operator interface and new graphical user interface pages were developed to more efficiently display system status data. A failure mode of the armor tile infrared thermometers (pyrometers), which serve to terminate beam pulsing if beam shine-through overheats wall thermal shielding inside the DIII-D tokamak, was characterized such that impending failures can be detected and repairs effected to mitigate beam system down-time. The hardware that controls gas flow to the beamline neutralizer cells was upgraded to reduce susceptibility to electromagnetic interference (EMI), and interlocking was provided to terminate beam pulsing in the event of insufficient neutralizer gas flow. Motivation, implementation, and results of these improvements are presented

  6. Extending DIII-D Neutral Beam Modulated Operations with a Camac Based Total on Time Interlock

    International Nuclear Information System (INIS)

    Baggest, D.S.; Broesch, J.D.; Phillips, J.C.

    1999-01-01

    A new total-on-time interlock has increased the operational time limits of the Neutral Beam systems at DIII-D. The interlock, called the Neutral Beam On-Time-Limiter (NBOTL), is a custom built CAMAC module utilizing a Xilinx 9572 Complex Programmable Logic Device (CPLD) as its primary circuit. The Neutral Beam Injection Systems are the primary source of auxiliary heating for DIII-D plasma discharges and contain eight sources capable of delivering 20MW of power. The delivered power is typically limited to 3.5 s per source to protect beam-line components, while a DIII-D plasma discharge usually exceeds 5 s. Implemented as a hardware interlock within the neutral beam power supplies, the NBOTL limits the beam injection time. With a continuing emphasis on modulated beam injections, the NBOTL guards against command faults and allows the beam injection to be safely spread over a longer plasma discharge time. The NBOTL design is an example of incorporating modern circuit design techniques (CPLD) within an established format (CAMAC). The CPLD is the heart of the NBOTL and contains 90% of the circuitry, including a loadable, 1 MHz, 28 bit, BCD count down timer, buffers, and CAMAC communication circuitry. This paper discusses the circuit design and implementation. Of particular interest is the melding of flexible modern programmable logic devices with the CAMAC format

  7. Customizable scientific web-portal for DIII-D nuclear fusion experiment

    International Nuclear Information System (INIS)

    Abla, G; Kim, E N; Schissel, D P

    2010-01-01

    Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

  8. Finding evidence for density fluctuation effects on electron cyclotron heating deposition profiles on DIII-D

    International Nuclear Information System (INIS)

    Brookman, M. W.; Austin, M. E.; Petty, C. C.

    2015-01-01

    Theoretical work, computation, and results from TCV [J. Decker “Effect of density fluctuations on ECCD in ITER and TCV,” EPJ Web of Conf. 32, 01016 (2012)] suggest that density fluctuations in the edge region of a tokamak plasma can cause broadening of the ECH deposition profile. In this paper, a GUI tool is presented which is used for analysis of ECH deposition as a first step towards looking for this broadening, which could explain effects seen in previous DIII-D ECH transport studies [K.W. Gentle “Electron energy transport inferences from modulated electron cyclotron heating in DIII-D,” Phys. Plasmas 13, 012311 (2006)]. By applying an FFT to the T e measurements from the University of Texas’s 40-channel ECE Radiometer, and using a simplified thermal transport equation, the flux surface extent of ECH deposition is determined. The Fourier method analysis is compared with a Break-In-Slope (BIS) analysis and predictions from the ray-tracing code TORAY. Examination of multiple Fourier harmonics and BIS fitting methods allow an estimation of modulated transport coefficients and thereby the true ECH deposition profile. Correlations between edge fluctuations and ECH deposition in legacy data are also explored as a step towards establishing a link between fluctuations and deposition broadening in DIII-D

  9. Model-based control of the resistive wall mode in DIII-D: A comparison study

    International Nuclear Information System (INIS)

    Dalessio, J.; Schuster, E.; Humphreys, D.A.; Walker, M.L.; In, Y.; Kim, J.-S.

    2009-01-01

    One of the major non-axisymmetric instabilities under study in the DIII-D tokamak is the resistive wall mode (RWM), a form of plasma kink instability whose growth rate is moderated by the influence of a resistive wall. One of the approaches for RWM stabilization, referred to as magnetic control, uses feedback control to produce magnetic fields opposing the moving field that accompanies the growth of the mode. These fields are generated by coils arranged around the tokamak. One problem with RWM control methods used in present experiments is that they predominantly use simple non-model-based proportional-derivative (PD) controllers requiring substantial derivative gain for stabilization, which implies a large response to noise and perturbations, leading to a requirement for high peak voltages and coil currents, usually leading to actuation saturation and instability. Motivated by this limitation, current efforts in DIII-D include the development of model-based RWM controllers. The General Atomics (GA)/Far-Tech DIII-D RWM model represents the plasma surface as a toroidal current sheet and characterizes the wall using an eigenmode approach. Optimal and robust controllers have been designed exploiting the availability of the RWM dynamic model. The controllers are tested through simulations, and results are compared to present non-model-based PD controllers. This comparison also makes use of the μ structured singular value as a measure of robust stability and performance of the closed-loop system.

  10. Non-linear instability of DIII-D to error fields

    International Nuclear Information System (INIS)

    La Haye, R.J.; Scoville, J.T.

    1991-10-01

    Otherwise stable DIII-D discharges can become nonlinearly unstable to locked modes and disrupt when subjected to resonant m = 2, n = 1 error field caused by irregular poloidal field coils, i.e. intrinsic field errors. Instability is observed in DIII-D when the magnitude of the radial component of the m = 2, n = 1 error field with respect to the toroidal field is B r21 /B T of about 1.7 x 10 -4 . The locked modes triggered by an external error field are aligned with the static error field and the plasma fluid rotation ceases as a result of the growth of the mode. The triggered locked modes are the precursors of the subsequent plasma disruption. The use of an ''n = 1 coil'' to partially cancel intrinsic errors, or to increase them, results in a significantly expanded, or reduced, stable operating parameter space. Precise error field measurements have allowed the design of an improved correction coil for DIII-D, the ''C-coil'', which could further cancel error fields and help to avoid disruptive locked modes. 6 refs., 4 figs

  11. An algorithm to provide real time neutral beam substitution in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Phillips, J.C.; Greene, K.L.; Hyatt, A.W.; McHarg, B.B. Jr.; Penaflor, B.G.

    1999-06-01

    A key component of the DIII-D tokamak fusion experiment is a flexible and easy to expand digital control system which actively controls a large number of parameters in real-time. These include plasma shape, position, density, and total stored energy. This system, known as the PCS (plasma control system), also has the ability to directly control auxiliary plasma heating systems, such as the 20 MW of neutral beams routinely used on DIII-D. This paper describes the implementation of a real-time algorithm allowing substitution of power from one neutral beam for another, given a fault in the originally scheduled beam. Previously, in the event of a fault in one of the neutral beams, the actual power profile for the shot might be deficient, resulting in a less useful or wasted shot. Using this new real-time algorithm, a stand by neutral beam may substitute within milliseconds for one which has faulted. Since single shots can have substantial value, this is an important advance to DIII-D's capabilities and utilization. Detailed results are presented, along with a description not only of the algorithm but of the simulation setup required to prove the algorithm without the costs normally associated with using physics operations time

  12. Customizable scientific web-portal for DIII-D nuclear fusion experiment

    Energy Technology Data Exchange (ETDEWEB)

    Abla, G; Kim, E N; Schissel, D P, E-mail: abla@fusion.gat.co [General Atomics, PO Box 85608, San Diego, California 92186-5608 (United States)

    2010-04-01

    Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

  13. Symplectic homoclinic tangles of the ideal separatrix of the DIII-D from type I ELMs

    Science.gov (United States)

    Punjabi, Alkesh; Ali, Halima

    2012-10-01

    The ideal separatrix of the divertor tokamaks is a degenerate manifold where both the stable and unstable manifolds coincide. Non-axisymmetric magnetic perturbations remove the degeneracy; and split the separatrix manifold. This creates an extremely complex topological structure, called homoclinic tangles. The unstable manifold intersects the stable manifold and creates alternating inner and outer lobes at successive homoclinic points. The Hamiltonian system must preserve the symplectic topological invariance, and this controls the size and radial extent of the lobes. Very recently, lobes near the X-point have been experimentally observed in MAST [A. Kirk et al, PRL 108, 255003 (2012)]. We have used the DIII-D map [A. Punjabi, NF 49, 115020 (2009)] to calculate symplectic homoclinic tangles of the ideal separatrix of the DIII-D from the type I ELMs represented by the peeling-ballooning modes (m,n)=(30,10)+(40,10). The DIII-D map is symplectic, accurate, and is in natural canonical coordinates which are invertible to physical coordinates [A. Punjabi and H. Ali, POP 15, 122502 (2008)]. To our knowledge, we are the first to symplectically calculate these tangles in physical space. Homoclinic tangles of separatrix can cause radial displacement of mobile passing electrons and create sheared radial electric fields and currents, resulting in radial flows, drifts, differential spinning, and reduction in turbulence, and other effects. This work is supported by the grants DE-FG02-01ER54624 and DE-FG02-04ER54793.

  14. Fast wave current drive experiment on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.; Harvey, R.W.; Lohr, J.; Luce, T.C.; Mayberry, M.J.; Prater, R.; Porkolab, M.; Baity, F.W.; Goulding, R.H.; Hoffman, J.D.; James, R.A.; Kawashima, H.

    1992-06-01

    One method of radio-frequency heating which shows theoretical promise for both heating and current drive in tokamak plasmas is the direct absorption by electrons of the fast Alfven wave (FW). Electrons can directly absorb fast waves via electron Landau damping and transit-time magnetic pumping when the resonance condition ω - κ parallele υ parallele = O is satisfied. Since the FW accelerates electrons traveling the same toroidal direction as the wave, plasma current can be generated non-inductively by launching FW which propagate in one toroidal direction. Fast wave current drive (FWCD) is considered an attractive means of sustaining the plasma current in reactor-grade tokamaks due to teh potentially high current drive efficiency achievable and excellent penetration of the wave power to the high temperature plasma core. Ongoing experiments on the DIII-D tokamak are aimed at a demonstration of FWCD in the ion cyclotron range of frequencies (ICRF). Using frequencies in the ICRF avoids the possibility of mode conversion between the fast and slow wave branches which characterized early tokamak FWCD experiments in the lower hybrid range of frequencies. Previously on DIII-D, efficient direct electron heating by FW was found using symmetric (non-current drive) antenna phasing. However, high FWCD efficiencies are not expected due to the relatively low electron temperatures (compared to a reactor) in DIII-D

  15. Software development on the DIII-D control and data acquisition computers

    International Nuclear Information System (INIS)

    Penaflor, B.G.; McHarg, B.B. Jr.; Piglowski, D.

    1997-11-01

    The various software systems developed for the DIII-D tokamak have played a highly visible and important role in tokamak operations and fusion research. Because of the heavy reliance on in-house developed software encompassing all aspects of operating the tokamak, much attention has been given to the careful design, development and maintenance of these software systems. Software systems responsible for tokamak control and monitoring, neutral beam injection, and data acquisition demand the highest level of reliability during plasma operations. These systems made up of hundreds of programs totaling thousands of lines of code have presented a wide variety of software design and development issues ranging from low level hardware communications, database management, and distributed process control, to man machine interfaces. The focus of this paper will be to describe how software is developed and managed for the DIII-D control and data acquisition computers. It will include an overview and status of software systems implemented for tokamak control, neutral beam control, and data acquisition. The issues and challenges faced developing and managing the large amounts of software in support of the dynamic and everchanging needs of the DIII-D experimental program will be addressed

  16. Comparison of discharges with core transport barriers on DIII-D and JET

    International Nuclear Information System (INIS)

    Luce, T.C.; Alper, B.; Challis, C.D.

    1997-07-01

    The basic phenomenology of discharges with core transport barriers is the same for DIII-D and JET. The limitations on performance in both cases are well described by MHD stability calculations. Since the discharge behavior of the two machines is so similar, it seems reasonable to apply a simple parameterization of fusion performance which describes well the best performance discharges on DIII-D. The highest fusion performance shot on JET has Q DD = 3.1 10 -3 at 3.2 MA. Scaling from the highest Q DD DIII-D single-null discharge would predict Q DD = 4.2 10 -3 for JET. Raising the plasma current to 4.0 MA would increase the projection to 6.6 10 -3 . Realization of such performance would require significant effort to develop lower q plasmas with an H-mode edge. Because the performance is so closely tied to the current profile, this class of discharges also shows significant potential for steady state if current profile control can be demonstrated

  17. Computational Study of Anomalous Transport in High Beta DIII-D Discharges with ITBs

    Science.gov (United States)

    Pankin, Alexei; Garofalo, Andrea; Grierson, Brian; Kritz, Arnold; Rafiq, Tariq

    2015-11-01

    The advanced tokamak scenarios require a large bootstrap current fraction and high β. These large values are often outside the range that occurs in ``conventional'' tokamak discharges. The GLF23, TGLF, and MMM transport models have been previously validated for discharges with parameters associated with ``conventional'' tokamak discharges. It has been demonstrated that the TGLF model under-predicts anomalous transport in high β DIII-D discharges [A.M. Garofalo et al. 2015 TTF Workshop]. In this research, the validity of MMM7.1 model [T. Rafiq et al. Phys. Plasmas 20 032506 (2013)] is tested for high β DIII-D discharges with low and high torque. In addition, the sensitivity of the anomalous transport to β is examined. It is shown that the MMM7.1 model over-predicts the anomalous transport in the DIII-D discharge 154406. In particular, a significant level of anomalous transport is found just outside the internal transport barrier. Differences in the anomalous transport predicted using TGLF and MMM7.1 are reviewed. Mechanisms for quenching of anomalous transport in the ITB regions of high-beta discharges are investigated. This research is supported by US Department of Energy.

  18. Modeling and experimental studies of the DIII-D neutral beam system

    Energy Technology Data Exchange (ETDEWEB)

    Crowley, B., E-mail: crowleyb@fusion.gat.com; Rauch, J.; Scoville, J.T.

    2015-10-15

    Highlights: • The issues surrounding proposals to increase neutral beam power are evaluated. • A tetrode version of the DIII-D ion source is modeled. • A neutralization efficiency of the DIII-D neutral beam is measured. • A power loading model of the neutral beam line is presented. - Abstract: In this paper, we present the results of beam physics experimental and modeling efforts aimed at learning from and building on the experience of the DIII-D off-axis neutral beam upgrade and other neutral beam system upgrades such as those at JET. The modeling effort includes electrostatic accelerator modeling (using a Poisson solver), gas dynamics modeling for the neutralizer and beam transport models for the beamline. Experimentally, spectroscopic and calorimetric techniques are used to evaluate the system performance. We seek to understand and ameliorate problems such as anomalous power deposition, originating from misdirected or excessively divergent beam particles, on a number of beamline components. We qualitatively and quantitatively evaluate possible project risks such as neutralization efficiency deficit and high voltage hold off associated with increasing the beam energy up to 105 keV.

  19. Progress towards a predictive model for pedestal height in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Leonard, A.W.; Snyder, P.B.; Osborne, T.H.; Petty, C.C.; Maggi, C.F.; Fenstermacher, M.E.; Owen, L.W.

    2009-01-01

    Recent DIII-D pedestal studies provide improved characterization of pedestal scaling for comparison with models. A new pedestal model accurately predicts the maximum achieved pedestal width and height in type I ELMing discharges over a large range of DIII-D operational space, including ITER demonstration discharges. The model is a combination of the peeling-ballooning theory for the MHD stability limits on the pedestal with a simple pedestal width scaling in which the width is proportional to the square root of the pedestal poloidal beta. Width scalings based on the ion toroidal or poloidal gyroradius are much poorer descriptions of DIII-D data. A mass scaling experiment in H and D provides support for a poloidal beta scaling and is not consistent with an ion poloidal gyroradius scaling. Studies of pedestal evolution during the inter-ELM cycle provide evidence that both the pedestal width and height increase during pedestal buildup. Model studies with a 1D kinetic neutrals calculation show that the temporal increase in density width cannot be explained in terms of increased neutral penetration depth. These studies show a correlation of pedestal width with both the square root of the pedestal poloidal beta and the square root of the pedestal ion temperature during the pedestal buildup.

  20. Customizable scientific web-portal for DIII-D nuclear fusion experiment

    Science.gov (United States)

    Abla, G.; Kim, E. N.; Schissel, D. P.

    2010-04-01

    Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

  1. Recent developments on the 110 GHz electron cyclotron installation on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Ponce, D.; Callis, R.W.; Cary, W.P.; Ferron, J.R.; Green, M.; Grunloh, H.J.; Gorelov, Y.; Lohr, J.; Ellis, R.A.

    2003-01-01

    Significant improvements are being implemented to the capability of the 110 GHz electron cyclotron system on the DIII-D tokamak. Chief among these is the addition of the fifth and sixth 1 MW class gyrotrons, increasing the power available for auxiliary heating and current drive by nearly 60%. These tubes use artificially grown diamond r.f. output windows to obtain high power with long pulse capability. The beams from these tubes are nearly Gaussian, facilitating coupling to the waveguide. A new fully articulating dual launcher capable of high speed spatial scanning has been designed and tested. The launcher has two axis independent steering for each waveguide. The mirrors can be rotated at up to 100 deg./s. A new feedback system linking the DIII-D Plasma Control System (PCS) with the gyrotron beam voltage waveform generators permits real-time feedback control of some plasma properties such as electron temperature. The PCS can use a variety of plasma monitors to generate its control signal, including electron cyclotron emission and Mirnov probes. Electron cyclotron heating and electron cyclotron current drive were used during this year's DIII-D experimental campaign to control electron temperature, density, and q profiles, induce an ELM-free H-mode, and suppress the m=2/n=1 neoclassical tearing mode. The new capabilities have expanded the role of EC systems in tokamak plasma control

  2. AORSA full wave calculations of helicon waves in DIII-D and ITER

    Science.gov (United States)

    Lau, C.; Jaeger, E. F.; Bertelli, N.; Berry, L. A.; Green, D. L.; Murakami, M.; Park, J. M.; Pinsker, R. I.; Prater, R.

    2018-06-01

    Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases. These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10%–20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.

  3. Data Analysis Software Tools for Enhanced Collaboration at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schachter, J.; Peng, Q.; Schissel, D.P.

    1999-01-01

    Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is GAP1otObj, an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is MDSplus, which is used at DIED as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are ReviewPlus and EFITviewer. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server

  4. Finding evidence for density fluctuation effects on electron cyclotron heating deposition profiles on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Brookman, M. W., E-mail: brookmanmw@fusion.gat.com; Austin, M. E. [Institute for Fusion Studies, University of Texas at Austin, MS 13-505, 3483 Dunhill St, San Diego, CA 92121-1200 (United States); Petty, C. C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

    2015-12-10

    Theoretical work, computation, and results from TCV [J. Decker “Effect of density fluctuations on ECCD in ITER and TCV,” EPJ Web of Conf. 32, 01016 (2012)] suggest that density fluctuations in the edge region of a tokamak plasma can cause broadening of the ECH deposition profile. In this paper, a GUI tool is presented which is used for analysis of ECH deposition as a first step towards looking for this broadening, which could explain effects seen in previous DIII-D ECH transport studies [K.W. Gentle “Electron energy transport inferences from modulated electron cyclotron heating in DIII-D,” Phys. Plasmas 13, 012311 (2006)]. By applying an FFT to the T{sub e} measurements from the University of Texas’s 40-channel ECE Radiometer, and using a simplified thermal transport equation, the flux surface extent of ECH deposition is determined. The Fourier method analysis is compared with a Break-In-Slope (BIS) analysis and predictions from the ray-tracing code TORAY. Examination of multiple Fourier harmonics and BIS fitting methods allow an estimation of modulated transport coefficients and thereby the true ECH deposition profile. Correlations between edge fluctuations and ECH deposition in legacy data are also explored as a step towards establishing a link between fluctuations and deposition broadening in DIII-D.

  5. Development of burning plasma and advanced scenarios in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.

    2005-01-01

    Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q ∼ 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque. (author)

  6. Lubricant coating of dowel for the ITER vacuum vessel gravity support

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.Y. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Ahn, H.J., E-mail: hjahn@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bak, J.S. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Choi, C.H.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Zauner, C. [KRP-Mechatec Engineering GbR, 85748 Garching b, Muenchen (Germany)

    2012-08-15

    The ITER vacuum vessel gravity supports located in the lower level shall sustain loads in radial, toroidal and vertical directions. The hinge type VVGS consists of two hinges, upper and lower blocks and dowels. In order to develop the design concept and verify the structural integrity of the hinge system, the design analysis has been performed in detail. Inclination of 15 Degree-Sign for the hinge based supporting system was introduced to provide centering force to make stable equilibrium state of the vacuum vessel. Due to this inclination the hinges are rotated by the radial expansion of the VV during operation and baking, respectively. If a dowel is seized in the hinge, the supporting system can be highly stressed due to the restrained displacement in the seized dowel. Therefore, solid lubricant coatings were suggested on dowels in order to avoid seizing in the sliding area. In this work, several sets of coupons were made with different coating materials to investigate the effect according to the selection of coating material. Also, a test facility was designed to cover the ITER relevant loading and boundary conditions, e.g. vacuum condition, temperature, contact pressure, cycles, etc. From those test results, the optimized coating method was found to avoid seizure of dowel in the ITER VVGS.

  7. Design and performance tests of gas circulation heating of JT-60U vacuum vessel

    International Nuclear Information System (INIS)

    Yotsuga, M.; Masuzaki, T.; Sago, H.; Nishikane, M.; Uchikawa, T.; Iritani, Y.; Murakami, T.; Horiike, H.; Neyatani, Y.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Miyachi, I.

    1992-01-01

    This paper reports that in the final stage of construction of the upgraded JT-60 device (JT-60U), baking tests of the vacuum vessel was performed. The vessel torus was heated-up to 300 degrees C by means of the nitrogen gas circulation system and electric heaters mounted on the outboard solid wall of the vessel. The design of the gas flow channels inside the double-wall structure of the vessel was done based on flow model tests, fluid analysis, and flow network analysis. The results of the baking tests were satisfactory. In maintaining 300 degrees C bake-out temperature, required heating power of the gas circulation system and outboard heaters was 520kW and 50kW, respectively. The temperature distribution over the vessel wall was within 300 ± 30 degrees C. It was also shown or suggested that heat-up and cool-down time is about 30 hours. The baking tests data have been reflected on operations for plasma experiments

  8. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  9. Analysis on ingress of coolant event in vacuum vessel using modified TRAC-BF1 code

    International Nuclear Information System (INIS)

    Ajima, Toshio; Kurihara, Ryoichi; Seki, Yasushi

    1999-08-01

    The Transient Reactor Analysis Code (TRAC-BF1) was modified on the basis of ICE experimental results so as to analyze the Ingress of Coolant Event (ICE) in the vacuum vessel of a nuclear fusion reactor. In the previous report, the TRAC-BF1 code, which was originally developed for the safety analysis of a light water reactor, had been modified for the ICE of the fusion reactor. And the addition of the flat structural plate model to the VESSEL component and arbitrary appointment of the gravity direction had been added in the TRAC-BF1 code. This TRAC-BF1 code was further modified. The flat structural plate model of the VESSEL component was enabled to divide in multi layers having different materials, and a part of the multi layers could take a buried heater into consideration. Moreover, the TRAC-BF1 code was modified to analyze under the low-pressure condition close to vacuum within range of the steam table. This paper describes additional functions of the modified TRAC-BF1 code, analytical evaluation using ICE experimental data and the ITER model with final design report (FDR) data. (author)

  10. Aluminium vacuum vessel/first surface conceptual design for a commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Culbert, M.

    1981-01-01

    The purpose of this investigation was to develop a design concept for a commercial tokamak hybrid reactor (CTHR) vacuum vessel/first surface system which satisfies the engineering requirements for a commercial environment. An important distinction between the design constraints associated with 'pure' fusion and fusion-fission hybrid power reactors is that energy extraction from the first wall is not critical from the point of view of hybrid system economics. This allows the consideration of low temperature structural material for first wall application. The mechanical arrangement consists of a series of internally finned aluminium tube banks running poloidally around the torus. The coolant manifolds are at the top and bottom of the torus. The vessel is divided into sectors, the length of which depends on the spacing between TF coils. The tubes in each sector are welded to tube sheets which are in turn welded to semi-cylindrical manifolds which distribute the coolant uniformly to the tubes. The tubes, which are approx. equal to 2.5 cm in diameter at the manifold location, traverse the torus poloidal periphery and change from a circular cross section to a 2:1 elliptical cross section at the horizontal midplane. The arched tube is designed to be self-supporting between the manifold locations. The vacuum vessel's first surface will be plasma flamed sprayed aluminum applied to the tubular wall. (orig./GG)

  11. Pulse discharge cleaning of the vacuum vessel of HL-1 tokamak

    International Nuclear Information System (INIS)

    Li Guodong; Zhu Yukun; Xiao Zhenggui; Sun Shouqi; Ze Mingrui

    1986-01-01

    The HL-1 Tokamak was test-operated on September 21, 1984. During the period of vacuum conditioning, including 60 hours of baking up to 200 deg C and 7 x 10 4 shots of pulse discharge cleaning, the calculated quantities of carbon and oxygen removed are equivalent to 24 and 6 monolayers, respectively. Then, 124 shots of tokamak discharge were performed with low level plasma parameters. The plasma current and pulse length achieved were 60 kA and 85 ms at the toroidal magnetic field of 15 kG. This paper described the techniques used and the effect on discharge characteristics of bakeout and pulse discharge cleaning of the vacuum vessel

  12. Structural analysis of TFTR vacuum vessel bellows and bellows cover sections

    International Nuclear Information System (INIS)

    Driesen, G.

    1975-10-01

    A structural evaluation of the bellows and bellows cover sections was undertaken in order to confirm the structural integrity of these TFTR vacuum vessel components in the prescribed operating environment. The evaluations investigate component stability, stress, and deflection behavior. The products of this investigation appearing in this report include; (1) Structural verification of the vacuum bellows as currently defined in an operating environment of one atmosphere external pressure and 93 0 C (200 0 F) uniform temperature. (2) The establishment of a structurally adequate design configuration for the bellows cover section. (3) The presentation of a parametric study which indicates the effects of varying some bellows cover section parameters in order to obtain acceptable variations of this design configuration. (4) A verification of bellows and bellows cover section integrity to preclude a fatigue type failure for reactor startup and shutdown cyclic life in the design environment

  13. Design of parallel intersector weld/cut robot for machining processes in ITER vacuum vessel

    International Nuclear Information System (INIS)

    Wu Huapeng; Handroos, Heikki; Kovanen, Janne; Rouvinen, Asko; Hannukainen, Petri; Saira, Tanja; Jones, Lawrence

    2003-01-01

    This paper presents a new parallel robot Penta-WH, which has five degrees of freedom driven by hydraulic cylinders. The manipulator has a large, singularity-free workspace and high stiffness and it acts as a transport device for welding, machining and inspection end-effectors inside the ITER vacuum vessel. The presented kinematic structure of a parallel robot is particularly suitable for the ITER environment. Analysis of the machining process for ITER, such as the machining methods and forces are given, and the kinematic analyses, such as workspace and force capacity are discussed

  14. Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

    International Nuclear Information System (INIS)

    Sannazzaro, G.; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H.

    2001-01-01

    In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses

  15. Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sannazzaro, G. E-mail: sannazg@itereu.de; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H

    2001-11-01

    In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses.

  16. Possible incorporation of a dee-shaped vacuum vessel in Doublet III

    International Nuclear Information System (INIS)

    Davis, L.; Rawls, J.M.

    1979-11-01

    The design of Doublet III allows relative straightforward incorporation of any of a number of possible dee-configuration vacuum vessels that can serve as relevant size tests of reactor regime devices. Configurations simulating those of JET, ETF and INTOR with plasma areas larger than TFTR can be attained with significant physics parameter results. Such modifications to Doublet III could be incorporated into planned upgrade activites with operations beginning in 1984, early enough to influence the designs of ETF and INTOR and test the scaling laws, poloidal coil system, and impurity control systems proposed for these ignition devices

  17. Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

    International Nuclear Information System (INIS)

    Onozuka, M.; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H.

    2001-01-01

    Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region

  18. Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H

    2001-11-01

    Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region.

  19. Mechanical impacts of poloidal eddy currents on the continuous vacuum vessel of a tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo.

    1996-11-01

    Poloidal eddy currents are induced on the continuous torus vacuum vessel by changes of the toroidal field during the machine start-up (toroidal field coil charge), shut-down (toroidal field coil discharge) and plasma disruption (plasma diamagnetism change). Analytic forms for the eddy currents flowing on the vessel, consequent pressures and forces acting on it are presented in this report. The results are applied to typical operation modes of the KT-2 tokamak. Stress analysis for two typical operation modes of toroidal field damping during a machine shut-gown and plasma energy quench during a plasma disruption were carried out using 3D FEM code (ANSYS 5.2). (author). 5 tabs., 22 figs., 9 refs

  20. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1993--September 30, 1994

    International Nuclear Information System (INIS)

    Lohr, J.

    1995-07-01

    The DIII-D tokamak research program is managed by General Atomics (GA) for the US Department of Energy (DOE). Major program participants include GA, Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Laboratory (ORNL), and the University of California together with several other national laboratories and universities. The DIII-D is a moderate sized tokamak with great flexibility and extremely capable subsystems. The primary goal of the DIII-D tokamak research program is to provide data for development of a conceptual physics blueprint for a commercially attractive fusion power plant. In so doing, the DIII-D program provides physics and technology R ampersand D output to aid the International Thermonuclear Experimental Reactor (ITER) and the Princeton Tokamak Physics Experiment (TPX) projects. Specific DIII-D objectives include the achievement of steady-state plasma current as well as the demonstration of techniques for radio frequency heating, divertor heat removal, particle exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion in plasmas with high beta and with high confinement. The long-range plan is organized with two principal elements, the development of an advanced divertor and the development of advanced tokamak concepts. These two elements have a common goal: an improved demonstration reactor (DEMO) with lower cost and smaller size than present DEMO concepts. In order to prepare for this long-range development, in FY94 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak studies, and Tokamak Physics

  1. Weld distortion prediction and control of the ITER vacuum vessel manufacturing mock-ups

    International Nuclear Information System (INIS)

    Ottolini, Marco; Barbensi, Andrea

    2014-01-01

    The fabrication of the ITER Vacuum Vessel Sectors is an unprecedented challenge, due to their dimensions, the close tolerances, the complex 'D' shape. The technological issues were faced by the production of full scale mock ups to confirm the manufacturing feasibility to achieve very tight tolerances and qualify the main manufacturing processes, by a step by step welding distortion control, by the qualification of not conventional NDT inspection techniques and by innovative 3D dimensional inspections. The Supplier is required to fabricate at least two mock ups, inboard and outboard, related to the manufacturing method of the VV Sectors, to demonstrate the control of the welding distortions to achieve tolerances, optimizing welding sequences and calibrating of welding distortions computer simulations. The stages of this preparatory activity are: prediction of welding distortion for fabrication mock ups representative of selected segments; demonstration that distortion predictions are consistent with experimental results from 3D dimensional inspection; understanding of reasons of possible deviations between numerical and experimental results and definition of action to solve these issues; demonstration that possible calculation simplifications, adopted to speed up the analysis process, do not affect significantly the welding distortion prediction. This paper describes the weld distortion prediction and control on the manufacturing mock-ups of ITER Vacuum Vessel Sectors, with particular emphasis to the lessons learned. (authors)

  2. Conditioning of SST-1 Tokamak Vacuum Vessel by Baking and Glow Discharge Cleaning

    International Nuclear Information System (INIS)

    Khan, Ziauddin; George, Siju; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Pathan, Firozkhan S.; Paravastu, Yuvakiran; Raval, Dilip C.; Babu, Gattu Ramesh; Khan, Mohammed Shoaib; Pradhan, Subrata

    2016-01-01

    Highlights: • SST-1 Tokamak was successfully commissioned. • Vacuum vessel was pumped down to 4.5 × 10"–"8 mbar after baking and continuous GDC. • GDC reduced the water vapour by additional 57% while oxygen was reduced by 50%. • Under this condition, an initial plasma breakdown with current of 40 kA for 75 ms was achieved. - Abstract: Steady-state Superconducting Tokamak (SST-1) vacuum vessel (VV) adopts moderate baking at 110 ± 10 °C and the limiters baking at 250 ± 10 °C for ∼ 200 h followed by glow discharge cleaning in hydrogen (GDC-H) with 0.15 A/m"2 current density towards its conditioning prior to plasma discharge experiment. The baking in SST-1 reduces the water (H_2O) vapor by 95% and oxygen (O_2) by 60% whereas the GDC reduces the water vapor by an additional 57% and oxygen by another 50% as measured with residual gas analyzer. The minimum breakdown voltage for H-GDC in SST-1 tokamak was experimentally observed to 300 V at 8 mbar cm. As a result of these adherences, SST-1 VV achieves an ultimate of 4.5 × 10"−"8 mbar with two turbo-molecular pumps with effective pumping speed of 3250 l/s. In the last campaign, SST-1 has achieved successful plasma breakdown, impurity burn through and a plasma current of ∼ 40 kA for 75 ms.

  3. Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Testoni, P., E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Albanese, R. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Lucca, F.; Roccella, M. [L.T. Calcoli S.a.S. Piazza Prinetti, 26/B, Merate, Lecco (Italy); Portone, A. [Fusion for Energy, Josep Plá n. 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Rubinacci, G. [Association Euratom/ENEA/CREATE, DIEL, Università Federico II di Napoli, Napoli 80125 (Italy); Ventre, S.; Villone, F. [Association Euratom/ENEA/CREATE, DAEIMI, Università di Cassino, Cassino 03043 (Italy)

    2013-10-15

    Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained.

  4. Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

    International Nuclear Information System (INIS)

    Testoni, P.; Albanese, R.; Lucca, F.; Roccella, M.; Portone, A.; Rubinacci, G.; Ventre, S.; Villone, F.

    2013-01-01

    Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained

  5. Progress and achievements of R and D activities for the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

    2001-01-01

    The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R and D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18 deg. toroidal sector, which is composed of two 9 deg. sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of ±3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R and D activities for the ITER vacuum vessel. (author)

  6. Progress and achievements of R&D activities for the ITER vacuum vessel

    Science.gov (United States)

    Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

    2001-04-01

    The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R&D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18° toroidal sector, which is composed of two 9° sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of +/-3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R&D activities for the ITER vacuum vessel.

  7. Manufacturing device for vacuum vessel of thermonuclear reactor and manufacturing method therefor

    International Nuclear Information System (INIS)

    Yanagi, Hiroshi; Shibui, Masanao; Uchida, Takaho

    1998-01-01

    The present invention provides a method of manufacturing a vacuum vessel of a thermonuclear reactor with no welding deformation. Namely, there are disposed a manufacturing device comprises a welding machine equipped with a plurality of welding torches which can conduct synchronizing welding and a torch positioning mechanism for positioning the plurality of welding torches each at an optional distance. Then, both ends of a splice plate can be welded by the plurality of welding torches under synchronization. Accordingly, joining portions of sectors of a vacuum vessel can be welded in the site with no deviation of beveling at joining portions between an outer wall and an inner wall with the splice plate due to welding deformation. In addition, the welding machine is mounted on a travelling type clamping mechanism stand or a travelling type clamping mechanism. With such a constitution, since the peripheries of the joining portions on the inner wall are clamped with each other by the travelling type clamping mechanism, no angular distortion is caused in any welded portion of the outer wall. (I.S.)

  8. An Approach for Selection of Flow Regime and Models for Conservative Evaluation of a Vessel Integrity Monitoring System for Water-Cooled Vacuum Vessels

    International Nuclear Information System (INIS)

    Pointer, W. David; Ruggles, Arthur E.

    2003-01-01

    Thin-walled vacuum containment vessels cooled by circulating water jackets are often utilized in research and industrial applications where isolation of equipment or experiments from the influences of the surrounding environment is desirable. The development of leaks in these vessels can result in costly downtime for the facility. A Vessel Integrity Monitoring System (VIMS) is developed to detect leak formation and estimate the size of the leak to allow evaluation of the risk associated with continued operation. A wide range of leak configurations and fluid flow phenomena are considered in the evaluation of the rate at which a tracer gas dissolved in the cooling jacket water is transported into the vacuum vessel. A methodology is presented that uses basic fluid flow models and careful evaluation of their ranges of applicability to provide a conservative estimate of the transport rates for the tracer gas and hence the time required for the VIMS to detect a leak of a given size

  9. The design study of the JT-60SU device. No. 4. The vacuum vessel and cryostat of JT-60SU

    International Nuclear Information System (INIS)

    Neyatani, Yuzuru; Ushigusa, Kenkichi; Tobita, Kenji

    1997-03-01

    The vacuum vessel and the cryostat for the JT-60 Super Upgrade (JT-60SU) have been designed. Two types of the complex materials for the vacuum vessel were chosen on the basis of the avoidance of tritium occlusion and the low irradiation, i.e. (1) SUS316 covered by tungsten plate (30mm thickness) as a γ-ray shielding, (2) Ti-6Al-4V alloy covered by SUS430 plate (1mm thickness) as a tritium protector. Selecting the double skin type of vacuum vessel with toroidally continued structure gave the basic design of the vacuum vessel satisfying the design criteria of the vessel strength for the electromagnetic force, heat load and the property of radiation shielding. The characteristics of the SUS316 covered by tungsten plate type is that as the tungsten can shield the γ-ray, the dose rate inside the vacuum vessel during the maintenance can reduce effectively. The advantage of the Ti-6Al-4V alloy covered by SUS430 plate type vacuum vessel is the quick reduction of the radioactive isotope because of no production of the isotopes with long half-life periods. Channel type and vertical type of the divertor were designed. The sector type of toroidally separated structure was selected for the remote handling. The material of the armor plate was not determined because no material endure the high heat load on the divertor. The cryostat composing the dome and the tank was designed. The electromagnetic force by the eddy current, generated at the plasma start up phase and at the quench of CS super-conducting coil, were small compared to the force produced by the stress limit. (author)

  10. Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Miyamoto, S.; Sugihara, M.; Shinya, K.; Nakamura, Y.; Toshimitsu, S.; Lukash, V.E.; Khayrutdinov, R.R.; Sugie, T.; Kusama, Y.; Yoshino, R.

    2012-01-01

    Highlights: ► Taking account of intervention of VS control, VDE simulations were carried out. ► Malfunctioning of VS circuit (positive feedback) enhances the vertical force. ► The worst case was explored for vertical force on the ITER vacuum vessel. ► We confirmed the force is still within the design margin even if the worst case. - Abstract: Vertical displacement events (VDEs) and disruptions usually take place under intervention of vertical stability (VS) control and the vertical electromagnetic force induced on vacuum vessels is potentially influenced. This paper presents assessment of the force that arises from the VS control in ITER VDEs using a numerical simulation code DINA. The focus is on a possible malfunctioning of the ex-vessel VS control circuit: radial magnetic field is unintentionally applied to the direction of enhancing the vertical displacement further. Since this type of failure usually causes the largest forces (or halo currents) observed in the present experiments, this situation must be properly accommodated in the design of the ITER vacuum vessel. DINA analysis shows that although the ex-vessel VS control modifies radial field, it does not affect plasma motion and current quench behavior including halo current generation because the vacuum vessel shields the field created by the ex-vessel coils. Nevertheless, the VS control modifies the force on the vessel by directly acting on the eddy current carried by the conducting structures of the vessel. Although the worst case was explored in a range of plasma inductance and pattern of VS control in combination with the in-vessel VS control circuit, the result confirmed that the force is still within the design margin.

  11. Calculation of the electromagnetic forces on the ASDEX upgrade vacuum vessel on disruption of the plasma current

    International Nuclear Information System (INIS)

    Preis, H.

    1986-01-01

    This study investigates the magnetic field diffusion through the vacuum vessel of the ASDEX Upgrade tokamak that occurs on sudden disruption of the plasma current. Eddy currents are thereby produced in the vessel wall. Their time behaviour and distribution are determined. Furthermore, the vessel is permeated by various magnetic fields which, together with the eddy currents, exert magnetic forces in the vessel wall. These are also calculated. These numerical analyses are performed for two of the modes of operation envisaged for ASDEX Upgrade: the so-called limiter and single-null magnetic field configurations. (orig.)

  12. The Resistive Wall Mode Feedback Control System on DIII-D

    International Nuclear Information System (INIS)

    J.T. Scoville; D.H. Kellman; S.G.E. Pronko; A. Nerem; R. Hatcher; D. O'Neill; G. Rossi; M. Bolha

    1999-01-01

    One of the primary instabilities limiting the performance of the plasma in advanced tokamak operating regimes is the resistive wall mode (RWM) [1]. The most common RWM seen in the DIII-D tokamak is originated by an n=1 ideal external kink mode which, in the presence of a resistive wall, is converted to a slowly growing RWM. The mode causes a reduction in plasma rotation, a loss of stored energy, and sometimes leads to plasma disruption. It routinely limits the performance of a tokamak operating near reactor relevant parameter levels. A system designed to actively control the RWM has recently been installed on the DIII-D tokamak for the control of low m n=1 modes. In initial experiments, the control system has been capable of delaying the onset of RWMs in energetic discharges for several hundred milliseconds. The feedback control system consists of detector coils connected via control software to high power current amplifiers driving the excitation coils. The three pairs of excitation coils are each driven by a current amplifier and a DC power supply. The control signal is derived from a set of six sensor coils that measure radial flux as low as one Gauss. The signals are digitally processed by realtime software in the DIII-D Plasma Control System (PCS) to create a command that is sent to the current amplifier, with a cycle time of approximately 100 micros. The amplifiers, designed and fabricated by Robicon Corporation to a specification developed by PPPL and GA, are bipolar devices capable of ±5 kA at 300 V, with an operating bandwidth of approximately 800 Hz (-3 dB)

  13. Increased power delivery from the DIII-D neutral beam injection system

    International Nuclear Information System (INIS)

    Colleraine, A.P.; Callis, R.W.; Hong, R.M.; Kellman, D.H.; Kim, J.; Langhorn, A.R.; Lee, R.; Phillips, J.C.; Wight, J.J.

    1989-12-01

    The neutral beam system installed on the DIII-D tokamak employs eight 80 kV Long Pulse Sources (LPS) mounted on four beamlines and was originally designed to deliver a nominal 12 MW of H degree power to a plasma for pulses of up to 5 sec duration. Lawrence Berkeley Laboratory designed the LPS for the US Fusion Program to fill the requirements of both the DIII-D and the TFTR machines. Essentially all source components are of a common design; the DIII-D version is therefore conservative in its rated parameters. Recently a neutron shield has been constructed around the torus hall allowing D degree injection to become routine. Because deuterium beams have a better neutralization efficiency, the nominal power delivery per source has been measured to be approximately 2 MW (for a total of 16 MW) without any modifications. However, by reoptimizing the voltage gradients in the source, the perveance can be increased without degrading the optics. A change of gradient grid voltage from 0.83 V accel to 0.79 V accel raises the perveance from 2.5 to 3.0 μPerv with a corresponding gain in beam power of about 20%. The arc power required also must be increased to the range of 100 to 120 kW but this is well within the design limits of the LPS. Further studies of our systems are now underway to assess the possibilities of raising V accel above 80 kV. An additional gain in power is possible by this technique. 6 refs., 6 figs

  14. The Bootstrap Current and Neutral Beam Current Drive in DIII-D

    International Nuclear Information System (INIS)

    Politzer, P.A.

    2005-01-01

    Noninductive current drive is an essential part of the implementation of the DIII-D Advanced Tokamak program. For an efficient steady-state tokamak reactor, the plasma must provide close to 100% bootstrap fraction (f bs ). For noninductive operation of DIII-D, current drive by injection of energetic neutral beams [neutral beam current drive (NBCD)] is also important. DIII-D experiments have reached ∼80% bootstrap current in stationary discharges without inductive current drive. The remaining current is ∼20% NBCD. This is achieved at β N [approximately equal to] β p > 3, but at relatively high q 95 (∼10). In lower q 95 Advanced Tokamak plasmas, f bs ∼ 0.6 has been reached in essentially noninductive plasmas. The phenomenology of high β p and β N plasmas without current control is being studied. These plasmas display a relaxation oscillation involving repetitive formation and collapse of an internal transport barrier. The frequency and severity of these events increase with increasing β, limiting the achievable average β and causing modulation of the total current as well as the pressure. Modeling of both bootstrap and NBCD currents is based on neoclassical theory. Measurements of the total bootstrap and NBCD current agree with calculations. A recent experiment based on the evolution of the transient voltage profile after an L-H transition shows that the more recent bootstrap current models accurately describe the plasma behavior. The profiles and the parametric dependences of the local neutral beam-driven current density have not yet been compared with theory

  15. Implementation of a quasi-realtime display of DIII-D neutral beam heating waveforms

    International Nuclear Information System (INIS)

    Phillips, J.C.

    1993-10-01

    The DIII-D neutral beam system employs eight 80 keV ion sources mounted on four beamlines to provide plasma heating to the DIII-D tokamak. The neutral beam system is capable of injecting over 20 MW of deuterium power with flexibility in terms of timing and modulation of the individual neutral beams. To maintain DIII-D's efficient tokamak shot cycle and make informed control decisions, it is important to be able to determine which beams fired, and exactly when, by the time the tokamak shot is over. Previously this information was available in centralized form only after a several minute wait. A cost-effective alternative to the traditional eight-channel storage oscilloscope has been implemented using off the shelf PC hardware and software. The system provides a real time display of injected neutral beam accelerator voltages and tokamak plasma current, as well an a summation waveform indicative of the total injected power as a function of time. The hardware consists of a Macintosh Centris 650 PC with a Motorola 68040 microprocessor. Data acquisition is accomplished using a National Instrument's 16-channel analog to digital conversion board for the Macintosh. The color displays and functionality were developed using National Instruments' LabView environment. Because the price of PCs has been decreasing rapidly and their capabilities increasing, this system is far less expensive than an eight-channel storage oscilloscope. As a flexible combination of PC and software, the system also provides much more capability than a dedicated oscilloscope, acting as the neutral beam coordinator's logbook, recording comments and availability statistics. Data such as shot number and neutral beam parameters are obtained over the local network from other computers and added to the display. Waveforms are easily archived to disk for future recall. Details of the implementation will be discussed along with samples of the displays and a description of the system's function and capabilities

  16. Pedestal width and ELM size identity studies in JET and DIII-D; implications for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Beurskens, M N A; Lomas, P; Saarelma, S; Balboa, I; Flanagan, J; Giroud, C; Kempenaars, M [EURATOM/UKAEA Fusion Association, Culham Sc. Centre, Abingdon, OX14 3DB (United Kingdom); Osborne, T H; Groebner, R; Leonard, A; Snyder, P B; Bray, B [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Horton, L D [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Frassinetti, L [Association EURATOM-VR, Alfven Laboratory, School of Electrical Engineering, KTH, Stockholm (Sweden); Nunes, I [Centro de Fusao Nuclear, Associacao EURATOM-IST, Lisboa (Portugal); Crombe, K [Department of Applied Physics, Ghent University, Rozier 44, 9000 Gent (Belgium); Giovannozzi, E [Associazione EURATOM-ENEA Sulla Fusione, Consorzio RFX Padova (Italy); Kohen, N [Association EURATOM-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Loarte, A [ITER Organization, CS 90 046, F-13067 Saint Paul lez Durance Cedex (France); Loennroth, J, E-mail: Marc.Beurskens@jet.u [Association EURATOM-Tekes, Helsinki University of Technology (Finland)

    2009-12-15

    The dependence of the H-mode edge transport barrier width on normalized ion gyroradius (rho* = rho/a) in discharges with type I ELMs was examined in experiments combining data for the JET and DIII-D tokamaks. The plasma configuration as well as the local normalized pressure (beta), collisionality (nu*), Mach number and the ratio of ion and electron temperature at the pedestal top were kept constant, while rho* was varied by a factor of four. The width of the steep gradient region of the electron temperature (T{sub e}) and density (n{sub e}) pedestals normalized to machine size showed no or only a weak trend with rho*. A rho{sup 1/2} or rho{sup 1} dependence of the pedestal width, given by some theoretical predictions, is not supported by the current experiments. This is encouraging for the pedestal scaling towards ITER as it operates at lower rho* than existing devices. Some differences in pedestal structure and ELM behaviour were, however, found between the devices; in the DIII-D discharges, the n{sub e} and T{sub e} pedestal were aligned at high rho* but the n{sub e} pedestal shifted outwards in radius relative to T{sub e} as rho* decreases, while on JET the profiles remained aligned while rho* was scanned by a factor of two. The energy loss at an ELM normalized to the pedestal energy increased from 10% to 40% as rho* increased by a factor of two in the DIII-D discharges but no such variation was observed in the case of JET. The measured pedestal pressures and widths were found to be consistent with the predictions from modelling based on peeling-ballooning stability theory, and are used to make projections towards ITER

  17. Upgrade of DIII-D toroidal magnetic field power supply controls

    International Nuclear Information System (INIS)

    Petrach, P.M.; Rouleau, A.R.; McNulty, R.D.; Patrick, D.B.; Walin, J.L.

    1993-11-01

    The toroidal magnetic field power supply for the DIII-D tokamak is of the 12 pulse line commutated variety. It consists of four individual modules and a main system control cabinet which are combined to deliver 127,000 A and 1000 V to the toroidal field coil. The modules are connected in a series-parallel configuration but can be run alone or two at a time as well. Normally on DIII-D experiments, the series-parallel configuration is required. The original design provided each individual module with its own voltage and current control loop and a main control loop. A problem with this design was that the individual control loops would cause a current sharing imbalance in the parallel modules if the calibrated loops drifted by the slightest amount. It was determined that individual control loops were not needed and a single phase lock firing circuit was employed in the system cabinet with fiber optic links to the modules for gate drive signals. Since all four modules have to be on line for DIII-D to operate, a problem in any of the five E ampersand I control loops resulted in the supply, and, therefore, the tokamak, being idled. By reducing the number of control loops to one, the sharing problem was eliminated, as well as 4 out of 5 potential control failures. The original supply employed relay logic for sequence control and fault monitoring. There were over 130 relays in each module plus an additional 100 in the system cabinet. The combination of the number of relays with the required interconnecting wiring, the age of the supply, the vibrations of the cabinets and the harsh environment, resulted in a continuously escalating number of phantom, and often intermittent, faults. The fault and sequence logic relays were replaced by a new Programmable Logic Controller (PLC). All existing interconnect wire was removed and replaced with multiconductor cables that connect directly from fault sensors and input devices to the PLC

  18. Implementation of a quasi-realtime display of DIII-D neutral beam heating waveforms

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, J.C.

    1993-10-01

    The DIII-D neutral beam system employs eight 80 keV ion sources mounted on four beamlines to provide plasma heating to the DIII-D tokamak. The neutral beam system is capable of injecting over 20 MW of deuterium power with flexibility in terms of timing and modulation of the individual neutral beams. To maintain DIII-D`s efficient tokamak shot cycle and make informed control decisions, it is important to be able to determine which beams fired, and exactly when, by the time the tokamak shot is over. Previously this information was available in centralized form only after a several minute wait. A cost-effective alternative to the traditional eight-channel storage oscilloscope has been implemented using off the shelf PC hardware and software. The system provides a real time display of injected neutral beam accelerator voltages and tokamak plasma current, as well an a summation waveform indicative of the total injected power as a function of time. The hardware consists of a Macintosh Centris 650 PC with a Motorola 68040 microprocessor. Data acquisition is accomplished using a National Instrument`s 16-channel analog to digital conversion board for the Macintosh. The color displays and functionality were developed using National Instruments` LabView environment. Because the price of PCs has been decreasing rapidly and their capabilities increasing, this system is far less expensive than an eight-channel storage oscilloscope. As a flexible combination of PC and software, the system also provides much more capability than a dedicated oscilloscope, acting as the neutral beam coordinator`s logbook, recording comments and availability statistics. Data such as shot number and neutral beam parameters are obtained over the local network from other computers and added to the display. Waveforms are easily archived to disk for future recall. Details of the implementation will be discussed along with samples of the displays and a description of the system`s function and capabilities.

  19. Location and repair of air leaks in the ATF vacuum vessel

    International Nuclear Information System (INIS)

    Schwenterly, S.W.; Gabbard, W.A.; Schaich, C.R.; Yarber, J.L.

    1989-01-01

    On the basis of partial pressure rate-of-rise and base pressure measurements, it was determined that the Advanced Toroidal Facility (ATF) vacuum vessel had an air leak in the low 10 -4 mbarx ell/s range. Pinpointing this leak by conventional helium leak-checking procedures was not possible, because large portions of the outside of the vessel are covered by the helical field coils and a structural shell. Various alternative leak detection schemes that were considered are summarized and their advantages and disadvantages noted. In the method ultimately employed, gum-rubber patches of various sizes ranging from 12.7 by 12.7 cm to 20.3 by 30.5 cm were positioned on the inside surfaces of the vessel and evacuated by the leak detector (LD). After roughly 5% of the surface was inspected in this way, a leak of >10 -5 mbar xL/s was discovered and localized to an area of 5 by 5 cm. Dye penetrant applied to this area disclosed three pinholes. Two small slag pockets were discovered while these points were being ground out. After these were rewelded, no further leakage could be found in the repaired area. Global leak rates measured after the machine was reevacuated indicated that this leak was about 30% of the overall leak rate. 1 ref., 5 figs., 1 tab

  20. Location and repair of air leaks in the ATF [Advanced Toroidal Facility] vacuum vessel

    International Nuclear Information System (INIS)

    Schwenterly, S.W.; Gabbard, W.A.; Schaich, C.R.; Yarber, J.L.

    1989-01-01

    On the basis of partial pressure rate-of-rise and base pressure measurements, it was determined that the Advanced Toroidal Facility (ATF) vacuum vessel had an air leak in the low 10 -4 mbar-ell/s range. Pinpointing this leak by conventional helium leak-checking procedures was not possible, because large portions of the outside of the vessel are covered by the helcial field coils and a structural shell. Various alternative leak-detection schemes that were considered are summarized and their advantages and disadvantages noted. In the method ultimately employed, gun-rubber patches of various sizes ranging from 12.7 by 12.7 cm to 20.3 by 30.5 cm were positioned on the inside surfaces of the vessel and evacuated by the leak detector (LD). After roughly 5% of the surface was inspected in this way, a leak of > 10 -5 mbar-ell/s was discovered and localized to an area of 5 by 5 cm. Dye penetrant applied to this area disclosed three pinholes. Two small slag pockets were discovered while these points were being ground out. After these were rewelded, no furthered leakage could be found in the repaired area. Global leak rates measured after the machine was reevacuated indicated that this leak was about 30% of the overall leak rate. 1 ref., 5 figs., 1 tab

  1. Experiments on ion cyclotron damping at the deuterium fourth harmonic in DIII-D

    International Nuclear Information System (INIS)

    Pinsker, R.I.; Petty, C.C.; Baity, F.W.; Bernabei, S.; Greenough, N.; Heidbrink, W.W.; Mau, T.K.; Porkolab, M.

    1999-05-01

    Absorption of fast Alfven waves by the energetic ions of an injected beam is evaluated in the DIII-D tokamak. Ion cyclotron resonance absorption at the fourth harmonic of the deuteron cyclotron frequency is observed with deuterium neutral beam injection (f = 60 MHz, B T = 1.9 T). Enhanced D-D neutron rates are evidence of absorption at the Doppler-shifted cyclotron resonance. Characteristics of global energy confinement provide further proof of substantial beam acceleration by the rf. In many cases, the accelerated deuterons cause temporary stabilization of the sawtooth (monster sawteeth), at relatively low rf power levels of ∼1 MW

  2. Experiments on ion cyclotron damping at the deuterium fourth harmonic in DIII-D

    International Nuclear Information System (INIS)

    Pinsker, R. I.; Baity, F. W.; Bernabei, S.; Greenough, N.; Heidbrink, W. W.; Mau, T. K.; Petty, C. C.; Porkolab, M.

    1999-01-01

    Absorption of fast Alfven waves by the energetic ions of an injected beam is evaluated in the DIII-D tokamak. Ion cyclotron resonance absorption at the fourth harmonic of the deuteron cyclotron frequency is observed with deuterium neutral beam injection (f=60 MHz, B T =1.9 T). Enhanced D-D neutron rates are evidence of absorption at the Doppler-shifted cyclotron resonance. Characteristics of global energy confinement provide further proof of substantial beam acceleration by the rf. In many cases, the accelerated deuterons cause temporary stabilization of the sawtooth (''monster sawteeth''), at relatively low rf power levels of ∼1 MW. (c) 1999 American Institute of Physics

  3. Scaling of divertor heat flux profile widths in DIII-D

    International Nuclear Information System (INIS)

    Lasnier, C.J.; Makowski, M.A.; Boedo, J.A.; Allen, S.L.; Brooks, N.H.; Hill, D.N.; Leonard, A.W.; Watkins, J.G.; West, W.P.

    2011-01-01

    New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

  4. Environmental Assessment for the proposed modification and continued operation of the DIII-D facility

    International Nuclear Information System (INIS)

    1995-07-01

    The EA evaluates the proposed action of modifying the DIII-D fusion facility and conducting related research activities at the GA San Diego site over 1995-1999 under DOE contract number DE-ACO3-89ER51114. The proposed action is need to advance magnetic fusion research for future generation fusion devices such as ITER and TPX. It was determined that the proposed action is not a major action significantly affecting the quality of the human environment according to NEPA; therefore a finding of no significant impact is made and an environmental impact statement is not required

  5. Low-Z impurity transport in DIII-D - observations and implications

    International Nuclear Information System (INIS)

    Wade, M.R.; Houlberg, W.A.; Baylor, L.R.; West, W.P.; Baker, D.R.

    2001-01-01

    Impurity transport studies on DIII-D have revealed transport phenomena that are qualitatively consistent with that expected from turbulence transport theory in some cases and neoclassical transport theory in other cases. The transport model proposed here, which assumes that the total impurity transport is a linear sum of turbulence-driven transport and neoclassical transport, is shown to reproduce many of these observed features. This transport model is then applied to burn condition calculations, revealing that profile effects associated with neoclassical transport have a large effect on the maximum allowable impurity fraction in machines based on achieving neoclassical transport levels

  6. An overview of the DIII-D long pulse neutral beam system

    International Nuclear Information System (INIS)

    Callis, R.W.; Colleraine, A.P.; Hong, R.M.; Langhorn, A.R.; Lee, R.L.; Kim, J.; Phillips, J.C.; Wight, J.J.

    1988-09-01

    The four beamlines on the DIII-D tokamak have been upgraded to long pulse operation with the addition of eight 80 kV, 80 A, 5 sec long pulse sources. The eight sources have proven to be very reliable and have performed well. Up to 12 MW of H 0 has been injected into a plasma. Inertially cooled beam absorbers have proven capable of handling multi-second pulses. General performance characteristics and some recent long-pulse physics results are presented. 12 refs., 7 figs

  7. An overview of the DIII-D long pulse neutral beam system

    International Nuclear Information System (INIS)

    Callis, R.W.; Colleraine, A.P.; Hong, R.-M.; Langhorn, A.R.; Lee, R.L.; Kim, J.; Phillips, J.C.; Wight, J.J.

    1989-01-01

    The four beamlines on the DIII-D tokamak have been upgraded to long pulse operation with the addition of eight 80 kV, 80 A, 5 sec long pulse sources. The eight sources have proven to be very reliable and have performed well. Up to 12 MW of H 0 has been injected into a plasma. Inertially cooled beam absorbers have proven capable of handling multi-second pulses. General performance characteristics and some recent long-pulse physics results are presented. (author). 12 refs.; 7 figs.; 1 tab

  8. Analysis of plasma coupling with the prototype DIII-D ICRF antenna

    International Nuclear Information System (INIS)

    Ryan, P.M.; Hoffman, D.J.; Bigelow, T.S.; Baity, F.W.; Gardner, W.L.; Mayberry, M.J.; Rothe, K.E.

    1988-01-01

    Coupling to plasma in the H-mode is essential to the success of future ignited machines such as CIT. To ascertain voltage and current requirements for high-power second harmonic heating (2 MW in a 35- by 50-cm port), coupling to the DIII-D tokamak with a prototype compact loop antenna has been measured. The results show good loading for L-mode and limiter plasmas, but coupling 2 MW to an H-mode plasma demands voltages and currents near the limit of present technology. We report the technological analysis and progress that allow coupling of these power densities. 5 refs., 4 figs

  9. DIII-D Edge Plasma, Disruptions, and Radiative Processes. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Boedo, J. A.; Luckhardt, S.C.; Moyer, R. A.

    2001-01-01

    The scientific goal of the UCSD-DIII-D Collaboration during this period was to understand the coupling of the core plasma to the plasma-facing components through the plasma boundary (edge and scrape-off layer). To achieve this goal, UCSD scientists studied the transport of particles, momentum, energy, and radiation from the plasma core to the plasma-facing components under normal (e.g., L-mode, H-mode, and ELMs), and off-normal (e.g., disruptions) operating conditions.

  10. ACHIEVING AND SUSTAINING STEADY-STATE ADVANCED TOKAMAK CONDITIONS ON DIII-D

    International Nuclear Information System (INIS)

    WADE, MR; MURAKAMI, M; BRENNAN, DP; CASPER, TA; FERRON, JR; GAROFALO, AM; GREENFIELD, CM; HYATT, AW; JAYAKUMAR, R; KINSEY, JE; LAHAYE, RJ; LAO, LL; LAZARUS, EA; LOHR, J; LUCE, TC; PETTY, CC; POLITZER, PA; PRATER, R; STRAIT, EJ; TURNBULL, AD; WATKINS, JG; WEST, WP

    2002-01-01

    Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼ 85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds

  11. DIII-D Edge Plasma, Disruptions, and Radiative Processes. Final Report

    International Nuclear Information System (INIS)

    Boedo, J. A.; Luckhardt, S.C.; Moyer, R. A.

    2001-01-01

    The scientific goal of the UCSD-DIII-D Collaboration during this period was to understand the coupling of the core plasma to the plasma-facing components through the plasma boundary (edge and scrape-off layer). To achieve this goal, UCSD scientists studied the transport of particles, momentum, energy, and radiation from the plasma core to the plasma-facing components under normal (e.g., L-mode, H-mode, and ELMs), and off-normal (e.g., disruptions) operating conditions

  12. Achieving and sustaining steady-state advanced tokamak conditions on DIII-D

    International Nuclear Information System (INIS)

    Wade, M.R.; Murakami, M.; Brennan, D.P.

    2003-01-01

    Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds. (author)

  13. Complete Suppression of the m=2/n=1 NTM Using ECCD on DIII-D

    International Nuclear Information System (INIS)

    Petty, C.C.; La Haye, R.J.; Luce, T.C.; Humphreys, D.A.; Lohr, J.; Prater, R.; Austin, M.E.; Harvey, R.W.; Wade, M.R.

    2003-01-01

    Complete suppression of the m=2/n=1 neoclassical tearing mode (NTM) is reported for the first time using electron cyclotron current drive (ECCD) to noninductively generate current at the radius of the island O-point. Experiments on the DIII-D tokamak show that the maximum shrinkage of the m=2/n=1 island amplitude occurs when the ECCD location coincides with the q=2 surface. Estimates of the ECCD radial profile width from the island shrinkage are consistent with ray tracing calculations but may allow for a factor-of-1.5 broadening from electron radial transport

  14. Environmental Assessment for the proposed modification and continued operation of the DIII-D facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The EA evaluates the proposed action of modifying the DIII-D fusion facility and conducting related research activities at the GA San Diego site over 1995-1999 under DOE contract number DE-ACO3-89ER51114. The proposed action is need to advance magnetic fusion research for future generation fusion devices such as ITER and TPX. It was determined that the proposed action is not a major action significantly affecting the quality of the human environment according to NEPA; therefore a finding of no significant impact is made and an environmental impact statement is not required.

  15. Analysis of plasma coupling with the prototype DIII-D ICRF antenna

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, P.M.; Hoffman, D.J.; Bigelow, T.S.; Baity, F.W.; Gardner, W.L.; Mayberry, M.J.; Rothe, K.E.

    1988-01-01

    Coupling to plasma in the H-mode is essential to the success of future ignited machines such as CIT. To ascertain voltage and current requirements for high-power second harmonic heating (2 MW in a 35- by 50-cm port), coupling to the DIII-D tokamak with a prototype compact loop antenna has been measured. The results show good loading for L-mode and limiter plasmas, but coupling 2 MW to an H-mode plasma demands voltages and currents near the limit of present technology. We report the technological analysis and progress that allow coupling of these power densities. 5 refs., 4 figs.

  16. Embedded calibration system for the DIII-D Langmuir probe analog fiber optic links

    International Nuclear Information System (INIS)

    Watkins, J. G.; Rajpal, R.; Mandaliya, H.; Watkins, M.; Boivin, R. L.

    2012-01-01

    This paper describes a generally applicable technique for simultaneously measuring offset and gain of 64 analog fiber optic data links used for the DIII-D fixed Langmuir probes by embedding a reference voltage waveform in the optical transmitted signal before every tokamak shot. The calibrated data channels allow calibration of the power supply control fiber optic links as well. The array of fiber optic links and the embedded calibration system described here makes possible the use of superior modern data acquisition electronics in the control room.

  17. Baking system for ports of experimental advanced super-conducting tokamak vacuum vessel and thermal stress analysis

    International Nuclear Information System (INIS)

    Cheng Yali; Bao Liman; Song Yuntao; Yao Damao

    2006-01-01

    The baking system of Experimental Advanced Super-Conducting Toakamk (EAST) vacuum vessel is necessary to obtain the baking temperature of 150 degree C. In order to define suitable alloy heaters and achieve their reasonable layouts, thermal analysis was carried out with ANSYS code. The analysis results indicate that the temperature distribution and thermal stress of most parts of EAST vacuum vessel ports are uniform, satisfied for the requirement, and are safe based on ASME criterion. Feasible idea on reducing the stress focus is also considered. (authors)

  18. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

    2000-01-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ∼50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed

  19. Kinematic analysis on rail development into vacuum vessel for ITER blanket maintenance

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Shibanuma, Kiyoshi

    2006-01-01

    The vehicle manipulator system for blanket maintenance is used as a main driving mechanism for rail development, and three driving mechanisms d1, d2 (or d2') and d3 are used as cycle sequence of the repeated operations for rail development. This repeated operation can develop the articulated rail into the vacuum vessel. The rail development scenario, kinematic analysis model for rail development without any driving mechanisms in the rail joints, equations defined the angular between two rail links, identification of rail link at repeated operation, numerical analysis results on rail deployment under the forced position control of l i+1 , new rail development scenario using two driving mechanisms d1 and d2''under one cycle sequence of the repeated operations, and rail development test are reported. (S.Y.)

  20. FW/Blanket and vacuum vessel for RTO/RC ITER

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M.

    2000-01-01

    The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste