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Sample records for advanced tokamak operation

  1. INTEGATED ADVANCED TOKAMAK OPERATION ON DIII-D

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

    2002-10-01

    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.

  3. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1998-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  4. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1996-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  5. Profile control of advanced tokamak plasmas in view of continuous operation

    Energy Technology Data Exchange (ETDEWEB)

    Mazon, D., E-mail: Didier.Mazon@cea.fr

    2015-07-15

    The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named ‘advanced scenarios’ are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated ‘bootstrap’ current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described.

  6. Profile control of advanced tokamak plasmas in view of continuous operation

    Science.gov (United States)

    Mazon, D.

    2015-07-01

    The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named 'advanced scenarios' are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated 'bootstrap' current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described.

  7. Measurement requirements for the advanced tokamak operation of a burning plasma experiment

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R L [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Casper, T [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); Young, K M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States)

    2004-05-01

    The optimization of a tokamak towards steady state and high performance has been the focus of advanced tokamak (AT) research for the past decade. A central theme of AT research line is plasma control: control of the plasma shape; of the profiles of current, pressure, and rotation; of transport; and of MHD stability. To optimize the performance, measurements of crucial parameters such as the current density and the plasma pressure are required with appropriate spatial coverage and resolution. In addition, measurements of other parameters will be necessary to develop a fundamental understanding of the complex nonlinear interactions amongst the current density profile, the pressure profile and transport (e.g. turbulence) in high {beta} AT plasmas. Present day experiments are providing physics insight into what a burning plasma experiment (BPX) will require as measurements. Recent research has focused on MHD stability aspects such as the neoclassical tearing mode and resistive wall mode stabilization and control of the current profile. However, in burning plasmas, new factors such as alpha particles, with their heating contribution and their relationship to transport barriers, will be increasingly important. The close relationship between measurements and active control, and the resultant impact on the requirements, will be discussed.

  8. Advances in multi-megawatt lower hybrid technology in support of steady-state tokamak operation

    Science.gov (United States)

    Delpech, L.; Achard, J.; Armitano, A.; Artaud, J. F.; Bae, Y. S.; Belo, J. H.; Berger-By, G.; Bouquey, F.; Cho, M. H.; Corbel, E.; Decker, J.; Do, H.; Dumont, R.; Ekedahl, A.; Garibaldi, P.; Goniche, M.; Guilhem, D.; Hillairet, J.; Hoang, G. T.; Kim, H. S.; Kim, J. H.; Kim, H.; Kwak, J. G.; Magne, R.; Mollard, P.; Na, Y. S.; Namkung, W.; Oh, Y. K.; Park, S.; Park, H.; Peysson, Y.; Poli, S.; Prou, M.; Samaille, F.; Yang, H. L.; The Tore Supra Team

    2014-10-01

    It has been demonstrated that lower hybrid current drive (LHCD) systems play a crucial role for steady-state tokamak operation, owing to their high current drive (CD) efficiency and hence their capability to reduce flux consumption. This paper describes the extensive technology programmes developed for the Tore Supra (France) and the KSTAR (Korea) tokamaks in order to bring continuous wave (CW) LHCD systems into operation. The Tore Supra LHCD generator at 3.7 GHz is fully CW compatible, with RF power PRF = 9.2 MW available at the generator to feed two actively water-cooled launchers. On Tore Supra, the most recent and novel passive active multijunction (PAM) launcher has sustained 2.7 MW (corresponding to its design value of 25 MW m-2 at the launcher mouth) for a 78 s flat-top discharge, with low reflected power even at large plasma-launcher gaps. The fully active multijunction (FAM) launcher has reached 3.8 MW of coupled power (24 MW m-2 at the launcher mouth) with the new TH2103C klystrons. By combining both the PAM and FAM launchers, 950 MJ of energy, using 5.2 MW of LHCD and 1 MW of ICRH (ion cyclotron resonance heating), was injected for 160 s in 2011. The 3.7 GHz CW LHCD system will be a key element within the W (for tungsten) environment in steady-state Tokamak (WEST) project, where the aim is to test ITER technologies for high heat flux components in relevant heat flux density and particle fluence conditions. On KSTAR, a 2 MW LHCD system operating at 5 GHz is under development. Recently the 5 GHz prototype klystron has reached 500 kW/600 s on a matched load, and studies are ongoing to design a PAM launcher. In addition to the studies of technology, a combination of ray-tracing and Fokker-Planck calculations have been performed to evaluate the driven current and the power deposition due to LH waves, and to optimize the N∥ spectrum for the future launcher design. Furthermore, an LHCD system at 5 GHz is being considered for a future upgrade of the ITER

  9. Advances in lower hybrid current drive for tokamak long pulse operation. Technology and physics

    International Nuclear Information System (INIS)

    The paper gives a picture of the present status and understanding of technology and physics of Lower Hybrid Current Drive for long pulse operation in tokamaks, including the development of continuous wave (CW) high power klystrons, and its evolutions towards ITER. 3.7 GH / 700 kW CW klystrons produced in series by Thales Electron Devices are now in operation on Tore Supra. First series of eight klystrons delivered more than 4 MW to sustain non-inductive plasmas during 50 s. Moreover, a prototype of 500 kW CW klystron operating at 5 GHz developed for KSTAR by Toshiba Electron Tubes and Devices, and foreseen for ITER, is able to produce RF output powers of 300 kW / 800 s and 450 kW / 20 s on matched load. The situation on wave coupling and antennas is reported, with the latest Tore Supra results of the new CW Passive-Active Multi-junction (PAM) launcher: the antenna concept foreseen for ITER. First experiments with the PAM antenna in Tore Supra have provided extremely encouraging results in terms of power handling and coupling. Relevant ITER power density of ∼25 MW/m2 (2.7 MW of power injected into the plasma) has been maintained over ∼80 s. In addition, LH power of 2.7 MW has been coupled at a plasma-antenna distance of 10 cm. (author)

  10. Advanced Control Scenario of High-Performance Steady-State Operation for JT-60 Superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

    H. Tamai; Y. Kamada; A. Sakasai; S. Ishida; G. Kurita; M. Matsukawa; K. Urata; S. Sakurai; K. Tsuchiya; A. Morioka; Y. M. Miura; K. Kizu

    2004-01-01

    Plasma control on high-βN steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-βN exceeding the free-boundary limit is investigated with the stabilising wall of reduced-activated ferritic steel and the active feedback control of the in-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet, advanced plasma control for steady-state high performance operation could be expected.

  11. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    HUMPHREYS,D.A; FERRON,J.R; JOHNSON,R.D; LEUER,J.A; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M

    2003-10-01

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which 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 iteration of the design-test loop to optimize performance.

  12. Steady State Advanced Tokamak (SSAT): The mission and the machine

    International Nuclear Information System (INIS)

    Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

  13. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  14. Tokamak advanced pump limiter experiments and analysis

    International Nuclear Information System (INIS)

    Experiments with pump limiter modules on several operating tokamaks establish such limiters as efficient collectors of particles and has demonstrated the importance of ballistic scattering as predicted theoretically. Plasma interaction with recycling neutral gas appears to become important as the plasma density increases and the effective ionization mean free path within the module decreases. In limiters with particle collection but without active internal pumping, the neutral gas pressure is found to vary nonlinearly with the edge plasma density at the highest densities studies. Both experiments and theory indicate that the energy spectrum of gas atoms in the pump ducting is non-thermal, consistent with the results of Monte Carlo neutral atom transport calculations. The distribution of plasma power over the front surface of such modules has been measured and appears to be consistent with the predictions of simple theory. Initial results from the latest experiment on the ISX-B tokamak with an actively pumped limiter module demonstrates that the core plasma density can be controlled with a pump limiter and that the scrape-off layer plasma can partially screen the core plasma from gas injection. The results from module pump limiter experiments and from the theory and design analysis of advanced pump limiters for reactors are used to suggest the major features of a definitive, axisymmetric, toroidal belt pump limiter experiment

  15. Stability of infernal and ballooning modes in advanced tokamak scenarios

    NARCIS (Netherlands)

    Holties, H. A.; Huysmans, G. T. A.; Goedbloed, J. P.; Kerner, W.; Parail, V.V.; Soldner, F. X.

    1996-01-01

    A numerical parameter study has been performed in order to find MHD stable operating regimes for advanced tokamak experiments In this study we have concentrated on internal modes. Ballooning stability and stability with respect to infernal modes are considered. The calculations confirm that pressure

  16. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  17. KTM Tokamak operation scenarios software infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, V.; Baystrukov, K.; Golobkov, YU.; Ovchinnikov, A.; Meaentsev, A.; Merkulov, S.; Lee, A. [National Research Tomsk Polytechnic University, Tomsk (Russian Federation); Tazhibayeva, I.; Shapovalov, G. [National Nuclear Center (NNC), Kurchatov (Kazakhstan)

    2014-10-15

    One of the largest problems for tokamak devices such as Kazakhstan Tokamak for Material Testing (KTM) is the operation scenarios' development and execution. Operation scenarios may be varied often, so a convenient hardware and software solution is required for scenario management and execution. Dozens of diagnostic and control subsystems with numerous configuration settings may be used in an experiment, so it is required to automate the subsystem configuration process to coordinate changes of the related settings and to prevent errors. Most of the diagnostic and control subsystems software at KTM was unified using an extra software layer, describing the hardware abstraction interface. The experiment sequence was described using a command language. The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.

  18. MHD stability of advanced tokamak scenarios

    International Nuclear Information System (INIS)

    Tokamak plasmas with a non-monotonic q-profile (current profile) and negative shear in the plasma centre have been associated with improved confinement and large pressure gradients in the region of negative shear. In JET, this regime, has been obtained with pellet injection (the PEP mode) and in DIII-D by ramping the plasma elongation. In JET, the phase of improved confinement is transient and usually ends in a collapse due to an MHD instability which leads to a redistribution of the current and a monotonic q-profile. The infernal mode, which is driven by a large pressure gradient in the region of low shear near the minimum in the q-profile, is the most likely candidate for the observed instability. To extend the transient phase to steady state, control of the shape of the current density profile is essential. The modelling of these advanced tokamak scenarios with a non-monotonic q-profile using non-inductive current drive of lower hybrid waves, fast waves, and neutral beams is discussed elsewhere. The aim is to find suitable initial states and to maintain MHD stability when the plasma β is built up. For this purpose, the robustness of the MHD stability of these configurations is studied with respect to changes in the position and in the depth of the minimum in q, and in the shape of the q and pressure profile. The classes of equilibria chosen for the analysis are based on the modelling of the current-drive schemes for advanced tokamak scenarios in JET. The toroidal ideal and resistive MHD stability code CASTOR is used for the stability calculations. (author) 7 refs., 4 figs

  19. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  20. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    International Nuclear Information System (INIS)

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method

  1. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Hwan [Department of Nanoscale Semiconductor Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Hong, Suk-Ho [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); National Fusion Research Institute (NFRI), Daejeon 305-333 (Korea, Republic of); Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook, E-mail: joykang@hanyang.ac.kr [Department of Electrical Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of)

    2015-12-15

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.

  2. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    Science.gov (United States)

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-12-01

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.

  3. The contribution to the energy balance and transport in an advanced-fuel tokamak reactor

    International Nuclear Information System (INIS)

    The influence of synchrotron radiation emission on the energy balance of an advanced-fuel (such as D-3He, or catalyzed-D) tokamak plasma is considered. It is shown that a region in the β-T space exists, where the fusion energy delivered to the plasma overcomes synchrotron and bremsstrahlung energy losses, and which could then allow for ignited operation. 1-Dimensional codes results are also presented, which illustrate the main features of radial transport in a ignited, D-3He tokamak plasma

  4. DIII-D Advanced Tokamak Research Overview

    International Nuclear Information System (INIS)

    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 βNH 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

  5. A CONCEPT FOR NEXT STEP ADVANCED TOKAMAK FUSION DEVICE

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    A concept is introduced for initiating the design study of a special class of tokamak,which has a magnetic confinement configuration intermediate between contemporary advanced tokamak and the recently established spherical torus (ST,also well known by the name "spherical tokamak").The leading design parameter in the present proposal is a dimensionless geometrical parameter, the machine aspect ratio A=R0/a0=2.0,where the parameters a0 and R0 denote,respectively,the plasma (equatorial) minor radius and the plasma major radius.The aim of this choice is to technologically and experimentally go beyond the aspect ratio frontier (R0/a0≈2.5) of present day tokamaks and enter a broad unexplored domain existing on the (a0,R0) parameter space in current international tokamak database,between the data region already moderately well covered by the advanced conventional tokamaks and the data region planned to be covered by STs.Plasma minor radius a0 has been chosen to be the second basic design parameter, and consequently,the plasma major radius R0 is regarded as a dependent design parameter.In the present concept,a nominal plasma minor radius a0=1.2m is adopted to be the principal design value,and smaller values of a0 can be used for auxiliary design purposes,to establish extensive database linkage with existing tokamaks.Plasma minor radius can also be adjusted by mechanical and/or electromagnetic means to smaller values during experiments,for making suitable data linkages to existing machines with higher aspect ratios and smaller plasma minor radii.The basic design parameters proposed enable the adaptation of several confinement techniques recently developed by STs,and thereby a specially arranged central-bore region inside the envisioned tokamak torus,with retrieved space in the direction of plasma minor radius,will be available for technological adjustments and maneuverings to facilitate implementation of engineering instrumentation and real time high

  6. OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    BURRELL,KH

    2002-11-01

    OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, the authors have made significant progress in developing the building blocks needed for AT operation: (1) the authors have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {le} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. They have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiation power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet

  7. Operation of cryostat vacuum vessel of HT-7 superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)]. E-mail: yangyu@ipp.ac.cn; Su, M. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2006-11-15

    The superconducting tokamak HT-7 has been in operation for over 10 years. The safe and reliable operation of its cryostat vacuum vessel, which contains the superconducting coils is essential for each experimental run since the superconducting toroidal field coils are contained inside the vessel. In this paper, the operation is reviewed with the emphasis on the analysis on anomalous pressure rises and the corresponding solutions. It is shown that under close monitoring and timely handling, the cryostat vacuum vessel could still satisfy the requirements of the experimental operation despite of the material aging. This provides guideline for vacuum operating of HT-7. The experiences should be valuable for other superconducting projects as well, including a whole superconducting tokamak under construction, EAST.

  8. Overview of recent experimental results from the DIII-D advanced tokamak program

    International Nuclear Information System (INIS)

    The D III-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: 1) We have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; 2) Using this rotational stabilization, we have achieved βNH89 ≥ 10 for 4 τE limited by the neoclassical tearing mode; 3) Using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased βT by 60%; 4) We have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; 5) We have made the first integrated AT demonstration discharges with current profile control using ECCD; 6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and 7) We have demonstrated stationary tokamak operation for 6.5 s (36 τE) at the same fusion gain parameter of βNH89/q952 ≅ 0.4 as ITER but at much higher q95 = 4.2. We have developed general improvements applicable to conventional and advanced tokamak operating modes: 1) We have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 τE) with constant density and constant radiated power; 2) We have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; 3) We have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. (author)

  9. A Steady State Tokamak Operation by Use of Magnetic Monopoles

    OpenAIRE

    Narihara, K.

    1991-01-01

    A steady state tokamak operation based on a magnetic monopole circuit is considered. Circulation of a chain of iron cubes which trap magnetic monopoles generates the needed loop voltage. The monopole circuit is enclosed by a series of solenoid coils in which the magnetic field is feedback controlled so that the force on the circuit balance against the mechanical friction. The driving power is supplied through the current sources of poloidal, ohmic and solenoid coils. The current drive efficie...

  10. Halo current diagnostic system of experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chen, D. L.; Shen, B.; Sun, Y.; Qian, J. P., E-mail: jpqian@ipp.ac.cn; Wang, Y.; Xiao, B. J. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031 (China); Granetz, R. S. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States)

    2015-10-15

    The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

  11. Edge localized mode physics and operational aspects in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Becoulet, M [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Huysmans, G [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Sarazin, Y [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Garbet, X [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Ghendrih, Ph [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Rimini, F [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Joffrin, E [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Litaudon, X [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Monier-Garbet, P [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Ane, J-M [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Thomas, P [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Grosman, A [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Parail, V [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Wilson, H [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Lomas, P [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Vries, P de[Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Zastrow, K-D [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Matthews, G F [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Lonnroth, J [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Gerasimov, S [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom)] [and others

    2003-12-01

    Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios. Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs. High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high beta{sub p} are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and beta{sub p}. The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U

  12. Operation of a tokamak reactor in the radiative improved mode

    Science.gov (United States)

    Morozov, D. Kh.; Mavrin, A. A.

    2016-03-01

    The operation of a nuclear fusion reactor has been simulated within a model based on experimental results obtained at the TEXTOR-94 tokamak and other facilities in which quasistationary regimes were achieved with long confinement times, high densities, and absence of the edge-localized mode. The radiative improved mode of confinement studied in detail at the TEXTOR-94 tokamak is the most interesting such regime. One of the most important problems of modern tokamaks is the problem of a very high thermal load on a divertor (or a limiter). This problem is quite easily solved in the radiative improved mode. Since a significant fraction of the thermal energy is reemitted by an impurity, the thermal loading is significantly reduced. As the energy confinement time τ E at high densities in the indicated mode is significantly larger than the time predicted by the scaling of ITERH-98P(y, 2), ignition can be achieved in a facility much smaller than the ITER facility at plasma temperatures below 20 keV. The revealed decrease in the degradation of the confinement time τ E with an increase in the introduced power has been analyzed.

  13. Status of and prospects for advanced Tokamak regimes from multi-machine comparisons

    International Nuclear Information System (INIS)

    In this series of 21 slides the author presents an assessment of the present fusion performance of the advanced tokamaks (AT) regimes for non-inductive operation. These AT regimes include data from ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U and Tore-Supra. Only data from both the 'hybrid' without necessarily an ITB (internal transport barrier) or the 'steady-state' scenario have been considered because these scenarios are the 2 candidates for the ITER non inductive current drive operation. A new operational diagram is proposed: the figure of merit for fusion performance and confinement H(ITER-89P).βN/q295 versus the bootstrap current fraction e1/2.βP. In this diagram there is a continuous progression from the 'inductive' to the 'hybrid' and 'steady-state' tokamak operating mode. The following range of performance: H(ITER-89P).βN/q295 ∼ 0.3-0.4 at βP ∼ 1, q95 ∼ 5, is expected for Q = 5 non inductive current drive operation for ITER. Fusion performances tend to decrease with the pulse duration, so extending the plasma performances achieved on a short time scale requires operating safely far from the operational limits. Other conclusions concerning the operating domain of dimensionless parameters such as Larmor radius, collisionality, Mach number and ratio of ion to electron temperature are also presented. (A.C.)

  14. Second regime tokamak operation at large aspect ratio

    International Nuclear Information System (INIS)

    The equilibrium, stability, and transport properties of large aspect ratio tokamaks operating at the second stability regime are described theoretically using numerical and analytical techniques. It has been shown that, at large aspect ratio, significant current profile control is possible with relatively modest amounts of neutral beam current drive, and the power needed to access and maintain the second regime operation is calculated to be about 3 MW using the results of an integrated 1.5D transport and stability code. An example second regime experiment has been described and the results are presented of extensive calculations illustrating several possible operating scenarios, external and internal model stability boundaries, and the experimental features needed to evaluate and test the high beta tokamak theories. The theory which describes the stabilizing effect of energetic particles during high beta operation was extended to finite aspect ratio. A key technical problem for application of this technique appears to be caused by ripple transport. Plasma rotation effects are found to be generally destabilizing and several other schemes for improved access to the second stability regime are discussed including ponderomotive stabilization of the plasma edge region and active feedback control

  15. OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    BURRELL,HK

    2002-11-01

    OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, they have made significant progress in developing the building blocks needed for AT operation: (1) they have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {ge} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. The authors have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiated power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet

  16. Advanced ICRF antenna design for R-TOKAMAK

    International Nuclear Information System (INIS)

    The advanced ICRF antennas designed for the R-TOKAMAK (a proposal in the Institute of Plasma Physics, Nagoya University) are described. They are a standard loop antenna and a panel heater antenna for fast wave heating, and a waveguide antenna for ion Bernstein wave heating. The standard loop antenna is made of Al-alloy and has a simple structure to install because of radioactivation by D-T neutrons. For a high power heating, a new type antenna called 'Panel heater antenna' is proposed, and it has a wide radiation area and is able to select a parallel wave number. The field pattern of the panel heater antenna is measured. The feasibility of the waveguide antenna is discussed for the ion Bernstein wave heating. The radiation from the aperture of the double ridge waveguide is experimentally estimated with a load simulating the plasma. (author)

  17. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    International Nuclear Information System (INIS)

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available

  18. Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meglicki, Z

    1995-09-19

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs.

  19. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    International Nuclear Information System (INIS)

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new

  20. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Lee, K. W.; Hwang, C. K.; Hong, B. G.; Hong, G. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-05-01

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new.

  1. Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Qu, Hao; Zhang, Tao; Han, Xiang; Wen, Fei; Zhang, Shoubiao; Kong, Defeng; Wang, Yumin; Gao, Yu; Huang, Canbin; Cai, Jianqing; Gao, Xiang, E-mail: xgao@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei, Anhui 230031 (China)

    2015-08-15

    An X-mode polarized V band (50 GHz–75 GHz) radial and poloidal correlation reflectometry is designed and installed on Experimental Advanced Superconducting Tokamak (EAST). Two frequency synthesizers (12 GHz–19 GHz) are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together for launching through one single pyramidal antenna. Two poloidally separated antennae are installed to receive the reflected waves from plasma. This reflectometry system can be used for radial and poloidal correlation measurement of the electron density fluctuation. In ohmically heated plasma, the radial correlation length is about 1.5 cm measured by the system. The poloidal correlation analysis provides a means to estimate the fluctuation velocity perpendicular to the main magnetic field. In the present paper, the distance between two poloidal probing points is calculated with ray-tracing code and the propagation time is deduced from cross-phase spectrum. Fluctuation velocity perpendicular to the main magnetic field in the core of ohmically heated plasma is about from −1 km/s to −3 km/s.

  2. Statistical analysis of first period of operation of FTU Tokamak; Analisi statistica del primo periodo di operazioni del Tokamak FTU

    Energy Technology Data Exchange (ETDEWEB)

    Crisanti, F.; Apruzzese, G.; Frigione, D.; Kroegler, H.; Lovisetto, L.; Mazzitelli, G.; Podda, S. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Energia

    1996-09-01

    On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted.

  3. Mechanisms of Extending Operation Regionin the HL-1M Tokamak

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Stable operating region in the HL-1M tokamak has been extended by means ofwall conditioning, core fuelling and current control techniques. The mechanisms of the extensionare analyzed in this paper. Lithiumization diminishes the impurities and hydrogen recycling tothe lowest level. After lithiumization a high density up to 7×1019 m-3 was obtained easily bystrong gas puffing with ordinary ohmic discharge alone. More attractively we found that metalLi-coating exhibited the effects of wall stabilization. The low qa limit with higher density wasextended by a factor of 1.5~2 in comparison with that for boronization, and 1.2 for siliconization.Siliconization not only extended stable operating region significantly by itself, but also provideda good target plasma for other experiments of raising density limit. Core fuelling schemes arefavourable especially for siliconized wall with a higher level of medium-Z impurity (Z=14).After siliconization the maximum density near to 1020 m-3 was achieved by a combination ofsupersonic molecule beam injection and multipellet injection. The new defined slope of Hugilllimit illustrating more clearly the situation under low qa and high ne discharges was created toindicate the new region extended by combining Ip ramp-up with core fuelling. The slope with alarge Murakami coefficient increased by a factor of 50~60 %.

  4. Advanced tokamak physics experiments on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, T.S. [General Atomics, San Diego, CA (United States)

    1998-12-01

    Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy confinement time ({tau}{sub E}) and the plasma pressure (or beta {beta}{sub T} = 2 {mu}{sub 0} < p > /B{sub T}{sup 2}) can be achieved in steady-state conditions with high self driven bootstrap current fraction. In addition, effective power exhaust and impurity and particle control is required. Significant progress has been made in experimentally achieving regimes having the required performance in all of these aspects as well as in developing a theoretical understanding of the underlying physics. The authors have extended the duration of high performance ELMing H-mode plasmas with {beta}{sub N} H{sub iop} {approximately} 10 for 5 {tau}{sub E} ({approximately}1 s) and have demonstrated that core transport barriers can be sustained for the entire 5-s neutral beam duration in L-mode plasmas. Recent DIII-D work has advanced the understanding of improved confinement and internal transport barriers in terms of E x B shear stabilization of micro turbulence. With the aim of current profile control in discharges with negative central magnetic shear, they have demonstrated off-axis electron cyclotron current drive for the first time in a tokamak, finding an efficiency above theoretical expectations. MHD stability has been improved through shape optimization, wall stabilization, and modification of the pressure and current density profiles. Heat flux reduction and improved impurity and particle control have been realized through edge/divertor radiation and understanding and utilization of forced scrape off layer flow and divertor baffling.

  5. ADX: a high field, high power density, advanced divertor and RF tokamak

    Science.gov (United States)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  6. Overview of recent experimental results from the DIII-D advanced tokamak program

    International Nuclear Information System (INIS)

    The goals of DIII-D Advanced Tokamak (AT) experiments are to investigate and optimize the upper limits of energy confinement and MHD stability in a tokamak plasma, and to simultaneously maximize the fraction of non-inductive current drive. Significant overall progress has been made in the past 2 years, as the performance figure of merit βN H89P of 9 has been achieved in ELMing H-mode for over 16 τE without sawteeth. We also operated at βN∼7 for over 35 τE or 3 τR, with the duration limited by hardware. Real-time feedback control of β (at 95% of the stability boundary), optimizing the plasma shape (e.g., δ, divertor strike- and X-point, double/single null balance), and particle control (ne/nGW∼0.3, ZeffN H89P of 7. The QDB regime has been obtained to date only with counter neutral beam injection. Further modification and control of internal transport barriers (ITBs) has also been demonstrated with impurity injection (broader barrier), pellets, and ECH (strong electron barrier). The new Divertor-2000, a key ingredient in all these discharges, provides effective density, impurity and heat flux control in the high-triangularity plasma shapes. Discharges at ne/nGW∼1.4 have been obtained with gas puffing by maintaining the edge pedestal pressure; this operation is easier with Divertor-2000. We are developing several other tools required for AT operation, including real-time feedback control of resistive wall modes (RWMs) with external coils, and control of neoclassical tearing modes (NTMs) with electron cyclotron current drive (ECCD). (author)

  7. Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

    Science.gov (United States)

    Ren, Q.

    2015-11-01

    Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA

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

    Energy Technology Data Exchange (ETDEWEB)

    Song Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Anhui, Hefei 230031 (China)]. E-mail: songyt@ipp.ac.cn; Yao Damao [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Anhui, Hefei 230031 (China); Wu Songata [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Anhui, Hefei 230031 (China); Weng Peide [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Anhui, Hefei 230031 (China)

    2006-02-15

    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 {sub 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.

  9. Steady state operation of tokamaks. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The first IAEA Technical Committee Meeting (TCM) on Steady State Operation of Tokamaks was organized to discuss the operations of present long-pulse tokamaks (TRIAM-1M, TORE SUPRA, MT-7, HT-7M, HL-1M) and the plans for future steady-state tokamaks such as SST-1, CIEL, and HT-7U. This meeting, held from 13-15 October 1998, was hosted by the Academia Sinica Institute of Plasma Physics (ASIPP), Hefei, China. Participants from China, France, India, Japan, the Russian Federation, and the IAEA participated in the meeting. There were 18 individual presentations plus general discussions on many topics, including superconducting magnet systems, cryogenics, plasma position control, non-inductive current drive, auxiliary heating, plasma-wall interactions, high heat flux components, particle control, and data acquisition

  10. Development of burning plasma and advanced scenarios in the DIII-D tokamak

    International Nuclear Information System (INIS)

    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)

  11. Modular coils and finite-β operation of a quasi-axially symmetric tokamak

    International Nuclear Information System (INIS)

    Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nuehrenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (MERKEL, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (HIRSHMAN, S.P., VAN RIJ, W.I., MERKEL, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak

  12. Advanced antenna system for Alfven wave plasma heating and current drive in TCABR tokamak

    International Nuclear Information System (INIS)

    An advanced antenna system that has been developed for investigation of Alfven wave plasma heating and current drive in the TCABR tokamak is described. The main goal was the development of such a system that could insure the excitation of travelling single helicity modes with predefined wave mode numbers M and N. The system consists of four similar modules with poloidal windings. The required spatial spectrum is formed by proper phasing of the RF feeding currents. The impedance matching of the antenna with the four-phase oscillator is accomplished by resonant circuits which form one assembly unit with the RF feeders. The characteristics of the antenna system design with respect to the antenna-plasma coupling and plasma wave excitation, for different phasing of the feeding currents, are summarised. The antenna complex impedance Z=ZR+ZI is calculated taking into account both the plasma response to resonant excitation of fast Alfven waves and the nonresonant excitation of vacuum magnetic fields in conducting shell. The matching of the RF generator with the antenna system during plasma heating is simulated numerically, modelling the plasma response with mutually coupled effective inductances with corresponding active ZR and reactive ZI impedances. The results of the numerical simulation of the RF system performance, including both the RF magnetic field spectrum analysis and the modeling of the RF generator operation with plasma load, are presented. (orig.)

  13. Absolute intensity calibration of the 32-channel heterodyne radiometer on experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.; Zhao, H. L.; Liu, Y., E-mail: liuyong@ipp.ac.cn; Li, E. Z.; Han, X.; Ti, A.; Hu, L. Q.; Zhang, X. D. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Domier, C. W.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California at Davis, Davis, California 95616 (United States)

    2014-09-15

    This paper presents the results of the in situ absolute intensity calibration for the 32-channel heterodyne radiometer on the experimental advanced superconducting tokamak. The hot/cold load method is adopted, and the coherent averaging technique is employed to improve the signal to noise ratio. Measured spectra and electron temperature profiles are compared with those from an independent calibrated Michelson interferometer, and there is a relatively good agreement between the results from the two different systems.

  14. Advanced Tokamak Regimes in Alcator C-Mod with Lower Hybrid Current Drive

    Science.gov (United States)

    Parker, R.; Bonoli, P.; Gwinn, D.; Hutchinson, I.; Porkolab, M.; Ramos, J.; Bernabei, S.; Hosea, J.; Wilson, R.

    1999-11-01

    Alcator C-Mod has been proposed as a test-bed for developing advanced tokamak scenarios owing to its strong shaping, relatively long pulse length capability at moderate field, e.g. t ~ L/R at B = 5T and T_eo ~ 7keV, and the availability of strong ICRF heating. We plan to exploit this capability by installing up to 4 MW RF power at 4.6 GHz for efficient off-axis current drive by lower hybrid waves. By launching LH waves with a grill whose n_xx spectrum can be dynamically controlled over the range 2 2. Such reversed or nearly zero shear regimes have already been proposed as the basis of an advanced tokamak burning-plasma experiment-ATBX (M. Porkolab et al, IAEA-CN-69/FTP/13, IAEA,Yokohama 1998.), and could provide the basis for a demonstration power reactor. Theoretical and experimental basis for this advanced tokamak research program on C-Mod, including design of the lower hybrid coupler, its spectrum and current drive capabilities will be presented.

  15. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  16. Evidence for Anomalous Effects on the Current Evolution in Tokamak Operating Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Casper, T; Jayakumar, R; Allen, S; Holcomb, C; Makowski, M; Pearlstein, L; Berk, H; Greenfield, C; Luce, T; Petty, C; Politzer, P; Wade, M; Murakami, M; Kessel, C

    2006-10-03

    Alternatives to the usual picture of advanced tokamak (AT) discharges are those that form when anomalous effects alter the plasma current and pressure profiles and those that achieve stationary characteristics through mechanisms so that a measure of desired AT features is maintained without external current-profile control. Regimes exhibiting these characteristics are those where the safety factor (q) evolves to a stationary profile with the on-axis and minimum q {approx} 1 and those with a deeply hollow current channel and high values of q. Operating scenarios with high fusion performance at low current and where the inductively driven current density achieves a stationary configuration with either small or non-existing sawteeth may enhance the neutron fluence per pulse on ITER and future burning plasmas. Hollow current profile discharges exhibit high confinement and a strong ''box-like'' internal transport barrier (ITB). We present results providing evidence for current profile formation and evolution exhibiting features consistent with anomalous effects or with self-organizing mechanisms. Determination of the underlying physical processes leading to these anomalous effects is important for scaling of current experiments for application in future burning plasmas.

  17. Dynamic simulation of a planar flexible boom for tokamak in-vessel operations

    International Nuclear Information System (INIS)

    In this paper we present a dynamic model for the analysis of the vibrations of a planar articulated flexible boom to be used for tokamak in-vessel maintenance operations. The peculiarity of the mechanical structure of the boom enables us to consider separately the oscillations in the horizontal and vertical planes so that two separate models can be constructed for describing these phenomena. The results of simulations based on booms like that proposed for NET in-vessel operations are presented. (orig.)

  18. Pioneering superconducting magnets in large tokamaks: evaluation after 16 years of operating experience in tore supra

    International Nuclear Information System (INIS)

    The toroidal field (TF) system of Tore Supra (TS) is superconducting. After 16 years of operation it is possible to give an overview of the experience gained on a large superconducting system integrated in a large Tokamak. Quantitative data will be given, about the TF system for the cryogenic system and for the magnet system as well, concerning the number of plasmas shots and the availability of the machine. The origin and the number of breakdowns or incidents will be described, with emphasis on cryogenics, to document repairs and changes on the system components. Concerning the behaviour during operation, the Fast Safety Discharges (FSD) in operation are of particular interest for the Tokamak operation, as they interrupt it on a significant time of the order of one hour. This aspect is particularly documented. The approach followed to decrease the number of these FSD will be reported and explained. The Tore Supra Tokamak was the first important meeting between Superconductivity and Plasma Physics on a large scale. Overall, despite the differences in design and size, the accumulated experience over 16 years of operation is a useful tool to prepare the manufacturing and the operation of the ITER magnets. (authors)

  19. EFFECT OF PROFILES AND SHAPE ON IDEAL STABILITY OF ADVANCED TOKAMAK EQUILIBRIA

    Energy Technology Data Exchange (ETDEWEB)

    MAKOWSKI,MA; CASPER,TA; FERRON,JR; TAYLOR,TS; TURNBULL,AD

    2003-08-01

    OAK-B135 The pressure profile and plasma shape, parameterized by elongation ({kappa}), triangularity ({delta}), and squareness ({zeta}), strongly influence stability. In this study, ideal stability of single null and symmetric, double-null, advanced tokamak (AT) configurations is examined. All the various shapes are bounded by a common envelope and can be realized in the DIII-D tokamak. The calculated AT equilibria are characterized by P{sub 0}/

    {approx} 2.0-4.5, weak negative central shear, high q{sub min} (> 2.0), high bootstrap fraction, an H-mode pedestal, and varying shape parameters. The pressure profile is modeled by various polynomials together with a hyperbolic tangent pedestal, consistent with experimental observations. Stability is calculated with the DCON code and the resulting stability boundary is corroborated by GATO runs.

  20. Effect of Profiles and Space on Ideal Stability of Advanced Tokamak Equilibria

    Energy Technology Data Exchange (ETDEWEB)

    Makowski, M A; Casper, T A; Ferron, J R; Taylor, T S; Turnbull, A D

    2003-07-07

    The pressure profile and plasma shape, parameterized by elongation ({kappa}), triangularity ({delta}), and squareness ({zeta}), strongly influence stability. In this study, ideal stability of single null and symmetric, double-null, advanced tokamak (AT) configurations is examined. All the various shapes are bounded by a common envelope and can be realized in the DIII-D tokamak. The calculated AT equilibria are characterized by P{sub 0}/{l_angle}P{r_brace} {approx} 2.0-4.5, weak negative central shear, high q{sub min} (>2.0), high bootstrap fraction, an H-mode pedestal, and varying shape parameters. The pressure profile is modeled by various polynomials together with a hyperbolic tangent pedestal, consistent with experimental observations. Stability is calculated with the DCON code and the resulting stability boundary is corroborated by GATO runs.

  1. Observation of Energetic Particle Driven Modes Relevant to Advanced Tokamak Regimes

    Energy Technology Data Exchange (ETDEWEB)

    R. Nazikian; B. Alper; H.L. Berk; D. Borba; C. Boswell; R.V. Budny; K.H. Burrell; C.Z. Cheng; E.J. Doyle; E. Edlund; R.J. Fonck; A. Fukuyama; N.N. Gorelenkov; C.M. Greenfield; D.J. Gupta; M. Ishikawa; R.J. Jayakumar; G.J. Kramer; Y. Kusama; R.J. La Haye; G.R. McKee; W.A. Peebles; S.D. Pinches; M. Porkolab; J. Rapp; T.L. Rhodes; S.E. Sharapov; K. Shinohara; J.A. Snipes; W.M. Solomon; E.J. Strait; M. Takechi; M.A. Van Zeeland; W.P. West; K.L. Wong; S. Wukitch; L. Zeng

    2004-10-21

    Measurements of high-frequency oscillations in JET [Joint European Torus], JT-60U, Alcator C-Mod, DIII-D, and TFTR [Tokamak Fusion Test Reactor] plasmas are contributing to a new understanding of fast ion-driven instabilities relevant to Advanced Tokamak (AT) regimes. A model based on the transition from a cylindrical-like frequency-chirping mode to the Toroidal Alfven Eigenmode (TAE) has successfully encompassed many of the characteristics seen in experiments. In a surprising development, the use of internal density fluctuation diagnostics has revealed many more modes than has been detected on edge magnetic probes. A corollary discovery is the observation of modes excited by fast particles traveling well below the Alfven velocity. These observations open up new opportunities for investigating a ''sea of Alfven Eigenmodes'' in present-scale experiments, and highlight the need for core fluctuation and fast ion measurements in a future burning-plasma experiment.

  2. 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.

  3. DIII-D research operations

    International Nuclear Information System (INIS)

    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

  4. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    International Nuclear Information System (INIS)

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m2 and a surface heat flux of 1 MW/m2. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO2 rods. The helium coolant pressure is 5 MPa, entering the module at 2970C and exiting at 5500C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter

  5. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A. (eds.)

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  6. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  7. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    International Nuclear Information System (INIS)

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost

  8. Characteristics of edge-localized modes in the experimental advanced superconducting tokamak (EAST)

    DEFF Research Database (Denmark)

    Jiang, M.; Xu, G.S.; Xiao, C.;

    2012-01-01

    Edge-localized modes (ELMs) are the focus of tokamak edge physics studies because the large heat loads associated with ELMs have great impact on the divertor design of future reactor-grade tokamaks such as ITER. In the experimental advanced superconducting tokamak (EAST), the first ELMy high...... confinement modes (H-modes) were obtained with 1 MW lower hybrid wave power in conjunction with wall conditioning by lithium (Li) evaporation and real-time Li powder injection. The ELMs in EAST at this heating power are mostly type-III ELMs. They were observed close to the H-mode threshold power and produced...... small energy dumps (1-2% of the stored energy). Type-III ELMs produced a time-averaged peak heat flux of about 2 MW m(-2) on the target plate, a value which is similar to 10 times larger than that of ELM-free phases. A few isolated and large type-I-like ELM events were also observed in EAST...

  9. Analysis of line integrated electron density using plasma position data on Korea Superconducting Tokamak Advanced Research

    International Nuclear Information System (INIS)

    A 280 GHz single-channel horizontal millimeter-wave interferometer system has been installed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This system has a triangular beam path that does not pass through the plasma axis due to geometrical constraints in the superconducting tokamak. The term line density on KSTAR has a different meaning from the line density of other tokamaks. To estimate the peak density and the mean density from the measured line density, information on the position of the plasma is needed. The information has been calculated from tangentially viewed visible images using the toroidal symmetry of the plasma. Interface definition language routines have been developed for this purpose. The calculated plasma position data correspond well to calculation results from magnetic analysis. With the position data and an estimated plasma profile, the peak density and the mean density have been obtained from the line density. From these results, changes of plasma density themselves can be separated from effects of the plasma movements, so they can give valuable information on the plasma status.

  10. Simulations of the operational control of a cryogenic plant for a superconducting burning-plasma tokamak

    CERN Document Server

    Mitchell, N

    2001-01-01

    In recent proposals for next generation superconducting tokamaks, such as the ITER project, the nuclear burning plasma is confined by magnetic fields generated from a large set (up to 100 GJ stored energy) of superconducting magnets. These magnets suffer heat loads in operation from thermal and nuclear radiation from the surrounding components and plasma as well as eddy currents and AC losses generated within the magnets, together with the heat conduction through supports and resistive heat generated at the current lead transitions to room temperature. The initial cryoplant for such a tokamak is expected to have a steady state capacity of up to about 85 kW at 4.5 K, comparable to the system installed for LHC at CERN. Experimental tokamaks are expected to operate at least initially in a pulsed mode with 20-30 short plasma pulses and plasma burn periods each day. A conventional cryoplant, consisting of a cold box and a set of primary heat exchangers, is ill-suited to such a mode of operation as the instantaneou...

  11. Steady state operation of tokamaks. Report on the IAEA technical committee meeting held at Hefei, China, 13-15 October 1998

    International Nuclear Information System (INIS)

    The first IAEA Technical Committee Meeting on Steady State Operation of Tokamaks was held in October 1998 in Hefei, China. This meeting marks the timely start of Technical Committee Meetings in an important area of tokamak research since several experiments are already yielding impressive results and several new experiments are under construction. Among the ongoing experiments interesting results were reported from the superconducting tokamaks TRIAM 1-M, Tore Supra, and HT-7 and from a conventional tokamak, HL-1M

  12. Design and characterization of a 32-channel heterodyne radiometer for electron cyclotron emission measurements on experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Han, X.; Liu, X.; Liu, Y., E-mail: liuyong@ipp.ac.cn; Li, E. Z.; Hu, L. Q.; Gao, X. [Institution of Plasma Physics, Chinese Academy of Sciences, P. O. Box 1126, Hefei, Anhui 230031 (China); Domier, C. W.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California, Davis, California 95616 (United States)

    2014-07-15

    A 32-channel heterodyne radiometer has been developed for the measurement of electron cyclotron emission (ECE) on the experimental advanced superconducting tokamak (EAST). This system collects X-mode ECE radiation spanning a frequency range of 104–168 GHz, where the frequency coverage corresponds to a full radial coverage for the case with a toroidal magnetic field of 2.3 T. The frequency range is equally spaced every 2 GHz from 105.1 to 167.1 GHz with an RF bandwidth of ∼500 MHz and the video bandwidth can be switched among 50, 100, 200, and 400 kHz. Design objectives and characterization of the system are presented in this paper. Preliminary results for plasma operation are also presented.

  13. Cryopump operations with the tokamak neutral-beam-injector prototype

    International Nuclear Information System (INIS)

    The various components of the cryosystem are briefly discussed. They are: cryopanels and heat loads, divertor valve, and vacuum pumping system. In addition, some operations of the system are described

  14. Fluid-particle hybrid simulation on the transports of plasma, recycling neutrals, and carbon impurities in the Korea Superconducting Tokamak Advanced Research divertor region

    Science.gov (United States)

    Kim, Deok-Kyu; Hong, Sang Hee

    2005-06-01

    A two-dimensional simulation modeling that has been performed in a self-consistent way for analysis on the fully coupled transports of plasma, recycling neutrals, and intrinsic carbon impurities in the divertor domain of tokamaks is presented. The numerical model coupling the three major species transports in the tokamak edge is based on a fluid-particle hybrid approach where the plasma is described as a single magnetohydrodynamic fluid while the neutrals and impurities are treated as kinetic particles using the Monte Carlo technique. This simulation code is applied to the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak [G. S. Lee, J. Kim, S. M. Hwang et al., Nucl. Fusion 40, 575 (2000)] to calculate the peak heat flux on the divertor plate and to explore the divertor plasma behavior depending on the upstream conditions in its base line operation mode for various values of input heating power and separatrix plasma density. The numerical modeling for the KSTAR tokamak shows that its full-powered operation is subject to the peak heat loads on the divertor plate exceeding an engineering limit, and reveals that the recycling zone is formed in front of the divertor by increasing plasma density and by reducing power flow into the scrape-off layer. Compared with other researchers' work, the present hybrid simulation more rigorously reproduces severe electron pressure losses along field lines by the presence of recycling zone accounting for the transitions between the sheath limited and the detached divertor regimes. The substantial profile changes in carbon impurity population and ionic composition also represent the key features of this divertor regime transition.

  15. Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

    Science.gov (United States)

    Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; Holcomb, C. T.; Lao, L. L.; McKee, G. R.; Meneghini, O.; Staebler, G. M.; Grierson, B. A.; Qian, J. P.; Solomon, W. M.; Turnbull, A. D.; Holland, C.; Guo, W. F.; Ding, S. Y.; Pan, C. K.; Xu, G. S.; Wan, B. N.

    2016-06-01

    Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN , despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.

  16. Korea Superconducting tokamak advanced research project - Development of heating system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byung Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The heating and current drive systems for KSTAR based on multiple technologies (neutral beam, ion cyclotron, lower hybrid and electron cyclotron) have been designed to provide heating and current drive capabilities as well as flexibility in the control of current density and pressure profiles needed to meet the mission and research objectives of the machine. They are designed to operate for long-pulse lengths of up to 300 s. The NBI system initially delivers 8 MW of neutral beam power to the plasma from one co-directed beam line and shall be upgraded to provide 20 MW of neutral beam power with two co-directed beam lines plus one counter-directed beam line. It will be capable of being reconfigured such that the source arrangement is changed from horizontal to vertical stacking, with 6 MW beam power to the plasmas per beam line, in order to facilitate profile control. The RF system initially delivers 6 MW of rf power to the plasma, using a single four-strap antenna mounted in a midplane port. The system will be upgraded to proved 12 MW of rf power through 2 adjacent ports. In the first phase, we completed the basic design of RF system and the system have the capabilities to be operationable for pulse length up to 300 sec and in the 25-60 MHz frequency range. Lower hybrid system initially provides 1.5 MW LH rf power to the plasma at 3.7 GHz through a horizontal port, which has a capability to be operated for pulse length up to 300 sec, and shall be upgraded to provide 4.5 MW of LH rf power to the plasma. In the first phase, we completed the basic design of LHCD system which incorporate the TPX-type launcher and independently phase-changeable transmission system for the fully phased coupler. The ECH system will deliver up to 0.5 MW of power to the plasma for up to 0.5 sec. In the first phase, we completed the basic design of ECH system which includes an 84 GHz gyrotron system, a transmission system, and a launcher. The basic design of the low loss transmission system

  17. Investigation of component failure rates for pulsed versus steady state tokamak operation

    International Nuclear Information System (INIS)

    This report presents component failure rate data sources applicable to magnetic fusion systems, and defines multiplicative factors to adjust these data for specific use on magnetic fusion experiment designs. The multipliers address both long pulse and steady state tokamak operation. Thermal fatigue and radiation damage are among the leading reasons for large multiplier values in pulsed operation applications. Field failure rate values for graphite protective tiles are presented, and beryllium tile failure rates in laboratory testing are also given. All of these data can be used for reliability studies, safety analyses, design tradeoff studies, and risk assessments

  18. Study of the DEF Feedback Control System in AC Operation of Superconducting Tokamak

    Science.gov (United States)

    Wang, Hua; Luo, Jiarong; Yuan, Qiping; Xu, Congdong

    2007-02-01

    AC operation with multiple full cycles has been successfully performed on the superconducting tokamak HT-7. In the experiment, it was discovered that the saturation of the transformer magnetic flux with DEF, a signal name, was one of key aspects that affected the AC operation. The conditions of DEF were examined through the DEF feedback control system. By controlling the working patterns of the subsystems, namely the poloidal field control system and density control system, it was guaranteed that DEF would remain in the non-saturated status.

  19. Study of the DEF Feedback Control System in AC Operation of Superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

    WANG Hua; LUO Jiarong; YUAN Qiping; XU Congdong

    2007-01-01

    AC operation with multiple full cycles has been successfully performed on the superconducting tokamak HT-7. In the experiment, it was discovered that the saturation of the transformer magnetic flux with DEF, a signal name, was one of key aspects that affected the AC operation. The conditions of DEF were examined through the DEF feedback control system. By controlling the working patterns of the subsystems, namely the poloidal field control system and density control system, it was guaranteed that DEF would remain in the non-saturated status.

  20. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  1. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D3He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  2. Advanced Transport Operating Systems Program

    Science.gov (United States)

    White, John J.

    1990-01-01

    NASA-Langley's Advanced Transport Operating Systems Program employs a heavily instrumented, B 737-100 as its Transport Systems Research Vehicle (TRSV). The TRSV has been used during the demonstration trials of the Time Reference Scanning Beam Microwave Landing System (TRSB MLS), the '4D flight-management' concept, ATC data links, and airborne windshear sensors. The credibility obtainable from successful flight test experiments is often a critical factor in the granting of substantial commitments for commercial implementation by the FAA and industry. In the case of the TRSB MLS, flight test demonstrations were decisive to its selection as the standard landing system by the ICAO.

  3. RF-driven tokamak reactor with sub-ignited, thermally stable operation

    International Nuclear Information System (INIS)

    A Radio-Frequency Driven Tokamak Reactor (RFDTR) can use RF-power, programmed by a delayed temperature measurement, to thermally stabilize a power equilibrium below ignition, and to drive a steady state current. We propose the parameters for such a device generating approx. = 1600 MW thermal power and operating with Q approx. = 40 (= power out/power in). A one temperature zero-dimensional model allows simple analytical formulation of the problem. The relevance of injected impurities for locating the equilibrium is discussed. We present the results of a one-dimensional (radial) code which includes the deposition of the supplementary power, and compare with our zero-dimensional model

  4. Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Liqing; Zhang, Jizong; Chen, Kaiyun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn; Hu, Liqun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn; Li, Erzhong; Lin, Shiyao; Shi, Tonghui; Duan, Yanmin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhu, Yubao [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)

    2015-12-15

    Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well.

  5. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    Energy Technology Data Exchange (ETDEWEB)

    Lampert, M. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); BME NTI, Budapest (Hungary); Anda, G.; Réfy, D.; Zoletnik, S. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); Czopf, A.; Erdei, G. [Department of Atomic Physics, BME IOP, Budapest (Hungary); Guszejnov, D.; Kovácsik, Á.; Pokol, G. I. [BME NTI, Budapest (Hungary); Nam, Y. U. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-07-15

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

  6. Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma

    International Nuclear Information System (INIS)

    Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well

  7. A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak

    Science.gov (United States)

    Ren, J.; Zuo, G. Z.; Hu, J. S.; Sun, Z.; Yang, Q. X.; Li, J. G.; Zakharov, L. E.; Xie, H.; Chen, Z. X.

    2015-02-01

    A program involving the extensive and systematic use of lithium (Li) as a "first," or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.

  8. Commercial feasibility of fusion power based on the tokamak concept

    International Nuclear Information System (INIS)

    The impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants is determined. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  9. Control of the Resistive Wall Mode in Advanced Tokamak Plasmas on DIII-D

    International Nuclear Information System (INIS)

    Resistive wall mode (RWM) instabilities are found to be a limiting factor in advanced tokamak (AT) regimes with low internal inductance. Even small amplitude modes can affect the rotation profile and the performance of these ELMing H-mode discharges. Although complete stabilization of the RWM by plasma rotation has not yet been observed, several discharges with increased beam momentum and power injection sustained good steady-state performance for record time extents. The first investigation of active feedback control of the RWM has shown promising results: the leakage of the radial magnetic flux through the resistive wall can be successfully controlled, and the duration of the high beta phase can be prolonged. The results provide a comparative test of several approaches to active feedback control, and are being used to benchmark the analysis and computational models of active control

  10. Advanced Space Surface Systems Operations

    Science.gov (United States)

    Huffaker, Zachary Lynn; Mueller, Robert P.

    2014-01-01

    The importance of advanced surface systems is becoming increasingly relevant in the modern age of space technology. Specifically, projects pursued by the Granular Mechanics and Regolith Operations (GMRO) Lab are unparalleled in the field of planetary resourcefulness. This internship opportunity involved projects that support properly utilizing natural resources from other celestial bodies. Beginning with the tele-robotic workstation, mechanical upgrades were necessary to consider for specific portions of the workstation consoles and successfully designed in concept. This would provide more means for innovation and creativity concerning advanced robotic operations. Project RASSOR is a regolith excavator robot whose primary objective is to mine, store, and dump regolith efficiently on other planetary surfaces. Mechanical adjustments were made to improve this robot's functionality, although there were some minor system changes left to perform before the opportunity ended. On the topic of excavator robots, the notes taken by the GMRO staff during the 2013 and 2014 Robotic Mining Competitions were effectively organized and analyzed for logistical purposes. Lessons learned from these annual competitions at Kennedy Space Center are greatly influential to the GMRO engineers and roboticists. Another project that GMRO staff support is Project Morpheus. Support for this project included successfully producing mathematical models of the eroded landing pad surface for the vertical testbed vehicle to predict a timeline for pad reparation. And finally, the last project this opportunity made contribution to was Project Neo, a project exterior to GMRO Lab projects, which focuses on rocket propulsion systems. Additions were successfully installed to the support structure of an original vertical testbed rocket engine, thus making progress towards futuristic test firings in which data will be analyzed by students affiliated with Rocket University. Each project will be explained in

  11. Operation of an ITER relevant inspection robot on Tore Supra tokamak

    International Nuclear Information System (INIS)

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. CEA has developed a multipurpose carrier able to realize deployments in the plasma vessel without breaking the Ultra High Vacuum (UHV) and temperature conditioning. A 6 years R and D programme was jointly conducted by CEA-LIST Interactive Robotics Unit and the Institute for Magnetic Fusion Research (IRFM) in order to demonstrate the feasibility and reliability of an in-vessel inspection robot relevant to ITER requirements. The Articulated Inspection Arm robot (AIA) is an 8-m long multilink carrier with a payload up to 10 kg operable between plasma under tokamak conditioning environment; its geometry allows a complete close inspection of Plasma Facing Components (PFCs) of the Tore Supra vessel. Different tools are being developed by CEA to be plugged at the front head of the carrier. The diagnostic presently in operation consists in a viewing system offering accurate visual inspection of PFCs. Leak detection of first wall based on helium sniffing and laser compact system for carbon co-deposited layers characterizations or treatments are also considered for demonstration. In April 2008, the AIA robot equipped with its vision diagnostic has realized a complete deployment into Tore Supra and the first closed inspection of the vessel under UHV conditions. During the upcoming experimental campaign, the same operation will be performed under relevant conditions (10-6 Pa and 120 deg. C) after a conditioning phase at 200 deg. C to avoid outgassing pollution of the chamber. This paper describes the different steps of the project development, robot capabilities with the present operations conducted on Tore Supra and future requirements for making the robot a tool for tokamak routine operation.

  12. Operation of an ITER relevant inspection robot on Tore Supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gargiulo, Laurent [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)], E-mail: laurent.gargiulo@cea.fr; Bayetti, Pascal; Bruno, Vincent; Hatchressian, Jean-Claude; Hernandez, Caroline; Houry, Michael [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Keller, Delphine [CEA, LIST, Service de Robotique Interactive, F-92265 Fontenay aux Roses (France); Martins, Jean-Pierre [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Measson, Yvan; Perrot, Yann [CEA, LIST, Service de Robotique Interactive, F-92265 Fontenay aux Roses (France); Samaille, Frank [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)

    2009-06-15

    Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. CEA has developed a multipurpose carrier able to realize deployments in the plasma vessel without breaking the Ultra High Vacuum (UHV) and temperature conditioning. A 6 years R and D programme was jointly conducted by CEA-LIST Interactive Robotics Unit and the Institute for Magnetic Fusion Research (IRFM) in order to demonstrate the feasibility and reliability of an in-vessel inspection robot relevant to ITER requirements. The Articulated Inspection Arm robot (AIA) is an 8-m long multilink carrier with a payload up to 10 kg operable between plasma under tokamak conditioning environment; its geometry allows a complete close inspection of Plasma Facing Components (PFCs) of the Tore Supra vessel. Different tools are being developed by CEA to be plugged at the front head of the carrier. The diagnostic presently in operation consists in a viewing system offering accurate visual inspection of PFCs. Leak detection of first wall based on helium sniffing and laser compact system for carbon co-deposited layers characterizations or treatments are also considered for demonstration. In April 2008, the AIA robot equipped with its vision diagnostic has realized a complete deployment into Tore Supra and the first closed inspection of the vessel under UHV conditions. During the upcoming experimental campaign, the same operation will be performed under relevant conditions (10{sup -6} Pa and 120 deg. C) after a conditioning phase at 200 deg. C to avoid outgassing pollution of the chamber. This paper describes the different steps of the project development, robot capabilities with the present operations conducted on Tore Supra and future requirements for making the robot a tool for tokamak routine operation.

  13. ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

  14. DT and DHe3 tokamak test reactor concepts using advanced, high field superconductors

    International Nuclear Information System (INIS)

    If practical high temperature superconducting ceramic magnets can be developed, there could be a significant impact on reactor design. Potential advantages include a simpler, more robust magnet design, the possibility of demountable superconducting toroidal field coils and reduced shielding requirements. The high temperature superconductors can also have very high critical fields and could provide super high field operation. This could substantially increase eta tau/sub E/ values, reduce β requirements, and improve prospects for ohmic heating to ignition. The combination of moderately high β and super high field could make DHe3 operation possible in a JET size tokamak. In this paper we discuss possibilities for test reactor designs using high temperature high field superconductors. An illustrative design has a field at the plasma of 15 T. This reduces the required β to less than 2% for DT operation. The required plasma current is 5 MA. For a reactor size of R0 = 3.4m and a = 0.6m, the neutron wall loading is 3.3 MW/m2 at β = 1.5% for DT operation and an equal amount of fusion power is produced at β = 10% for DHe3 operation. One possible mode of operation is to use ohmic heating to ignition in a DT plasma followed by thermal runaway to DHe3 temperatures. 7 refs., 1 fig., 2 tabs

  15. Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas

    Science.gov (United States)

    Ribeiro, Celso

    2015-11-01

    The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.

  16. Status of tokamak research

    International Nuclear Information System (INIS)

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  17. Researches on operating region of Tokamak device with soft X-ray tomography

    Institute of Scientific and Technical Information of China (English)

    李林忠; 梁荣庆; 尹协锦; 邱励俭

    1997-01-01

    The structures of three operating regions in HT-6B Tokamak have been studied by soft X-ray tomo-graphic system with high sensibility and high time-space resolution. One of the requisites for forming sawtooth discharge is the effective heating action in the central region. In the sawtooth region there are five evolutional phases and five types of magnetic surface structures correspondingly; that is, the concentric, the eccentric, the double-core, the "MHD-type" and the "ultra-MHD type" magnetic surface structures. In the MHD oscillation region, there is a stable "MHD-type" magnetic surface structure. It consists of a crescent "hot core" and a circular "cold bubble" and rotates in the diamagnetic direction of electrons. In the resonant region, the resonant helical field improves the heating status and suppresses the MHD disturbances; therefore the single "MHD-type" magnetic surface changes into a sawtooth type one

  18. Estimation of Charge Exchange Recombination Emission Based on Diagnostic Neutral Beam on the Experimental Advanced Superconducting Tokamak

    Institute of Scientific and Technical Information of China (English)

    ZHANG Xian-Mei; WAN Bao-Nian; WU Zhen-Wei

    2007-01-01

    Diagnostic neutral beam (DNB) attenuation and charge exchange recombination emission are estimated on EAST tokamak. Approximately 40% of the beam with the energy of 50 keV can reach the plasma centre (r = 0) for the typical parameters of the Experimental Advanced Superconducting Tokamak (EAST) plasma. Emissivities of CVI (n = 8 → 7, 529.0nm) and OVⅢ (n = 10 → 9, 607.0 nm) visible charge exchange recombination emissions based on the DNB are estimated. The emissivities of the visible bremsstrahlung emission near this wavelength are also calculated for comparison. The results show that the charge exchange recombination emission is about two orders of magnitude greater than the bremsstrahlung emission. It is theoretically indicated that the ratio of signal of charge exchange recombination spectroscopy to the noise from background bremsstrahlung emission,S/N, is large enough in the EAST tokamak with the typical designed parameters. The present results are helpful for experiment design of charge-exchange recombination spectroscopy based on the DNB in the EAST tokamak.

  19. Startup scenarios of an advanced fuel tokamak: First wall and shield thermal response

    International Nuclear Information System (INIS)

    Three different startup scenarios, one using pure D-3He, He, one using pure D-T to assist reaching the D-3 He operating point, and one using a mixture of D-T-3 He, have been analyzed, for the startup of ARIES-III. ARIES-III is a conceptual D-3He tokamak fusion power reactor operating in a second stability configuration. The process of starting the plasma up and bringing it to the desired operating point has been optimized to minimize the need for auxiliary ICRF heating during startup. In the second and third startup scenarios, seeding the plasma with tritium during startup reduces the amount of ICRF power required, but leads to a 14 MeV neutron pulse. Neutronics calculations have been performed to generate the nuclear heating profiles in the first wall and shield. The neutronics results were scaled with the neutron power to determine the nuclear heating profiles at different times during the startup phase. In this work, a two-dimensional transient thermal analysis is performed for the startup phases and the temperature distribution in the first wall and shield as a function of time is presented. The analysis is performed for the worst conditions at the midplane of the outboard region

  20. Numerical analysis of high Mach flow and flow reversal in the experimental advanced superconducting tokamak divertor

    Institute of Scientific and Technical Information of China (English)

    Ou Jing; Yang Jin-Hong

    2011-01-01

    The B2-Eirene (SOLPS 4.0) code package is used to investigate the plasma parallel flow,i.e.,the scrape-off layer (SOL) flow,in the experimental advanced superconducting tokamak (EAST) divertor. Simulation results show that the SOL flow in the divertor region can exhibit complex behaviour,such as a high Mach flow and flow reversal in different plasma regimes. When the divertor plasma is in the detachment state,the high Mach flow with approaching or exceeding sonic speed is observed away from the target plate in our simulation. When the divertor plasma is in the high recycling The driving mechanisms for the high Mach flow and the reversed flow are analysed theoretically through momentum and continuity equations,respectively. The profile of the ionization sources is shown to be a possible formation condition causing the complex behaviour of the SOL flow. In addition,the effects of the high Mach flow and the flow reversal on the impurity transport are also discussed in this paper.

  1. Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

    Science.gov (United States)

    Li, Y. L.; Xu, G. S.; Tritz, K.; Zhu, Y. B.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L.

    2015-12-01

    A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.

  2. PROGRESS TOWARD SUSTAINED HIGH-PERFORMANCE ADVANCED TOKAMAK DISCHARGES IN DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    J.R. FERRON; D.P. BREEAN; T.A. CASPER; A.M. GAROFALO; C.M. GREENFIELD; A.W. HYATT; R.J. JAYAKUMAR; L.C. JOHNSON; J.E. KINSEY; R.J. LaHAYE; L.L. LAO; E.A. LAZARUS; J. LOHR; T.C. LUCE; M. MURAKAMI; M. OKABAYASHI; C.C. PETTY; P.A. POLITZER; R. PRATER; H. REIMERDES; E.J. STRAIT; T.S. TAYLOR; A.D. TURNBULL; J.G. WATKINS; M.R. WADE; W.P. WEST

    2002-07-01

    Key elements of a sustained advanced tokamak discharge in DIII-D are a large fraction of the total current from bootstrap current (f{sub BS}) and parameters that optimize the capability to use electron cyclotron current drive (ECCD) at {rho} {approx} 0.5 to maintain the desired current profile [1-4]. Increased f{sub BS} results from increasing both the normalized beta ({beta}{sub N}) and the minimum value of the safety factor (q{sub min}). Off-axis ECCD is, for the available gyrotron power, optimized at high {beta}{sub N}, high electron temperature (T{sub e}) and low electron density (n{sub e}). As previously reported [2-4], these required elements have been separately demonstrated: density control at high {beta}{sub N} with n{sub e} {le} 5 x 10{sup 19} m{sup -3} using divertor-region pumping, stability at high {beta}, and off-axis ECCD at the theoretically predicted efficiency. This report summarizes recent work on optimizing and integrating these results through evaluation of the dependence of the beta limit on q{sub min} and q{sub 95}, exploration of discharges with relatively high q{sub min}, testing of feedback control of T{sub e} for control of the q profile evolution, and modification of the current profile time evolution when ECCD is applied.

  3. Characterisation, modelling and control of advanced scenarios in the european tokamak jet

    International Nuclear Information System (INIS)

    The advanced scenarios, developed for less than ten years with the internal transport barriers and the control of current profile, give rise to a 'new deal' for the tokamak as a future thermonuclear controlled fusion reactor. The Joint European Torus (JET) in United Kingdom is presently the most powerful device in terms of fusion power and it has allowed to acquire a great experience in these improved confinement regimes. The reduction of turbulent transport, considered now as closely linked to the shape of current profile optimised for instance by lower hybrid current drive or the self-generated bootstrap current, can be characterised by a dimensionless criterion. Most of useful information related to the transport barriers are thus available. Large database analysis and real time plasma control are envisaged as attractive applications. The so-called 'S'-shaped transport models exhibit some interesting properties in fair agreement with the experiments, while the non-linear multivariate dependencies of thermal diffusivity can be approximated by a neural network, suggesting a new approach for transport investigation and modelling. Finally, the first experimental demonstrations of real time control of internal transport barriers and current profile have been performed on JET. Sophisticated feedback algorithms have been proposed and are being numerically tested to achieve steady-state and efficient plasmas. (author)

  4. Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Tritz, K. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Zhu, Y. B. [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)

    2015-12-15

    A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.

  5. Edge plasma issues of the tokamak FAST (Fusion Advanced Studies Torus) in reactor relevant conditions

    International Nuclear Information System (INIS)

    Among the R and D missions for possible new European plasma fusion devices, the FAST project will address the issue of 'First wall materials and compatibility with ITER /DEMO relevant plasmas'. FAST can operate with ITER relevant values of P/R (up to 22 MW/m, against the ITER 24 MW/m, inclusive of the alpha particles power), thanks to its compactness; thus it can investigate the physics of large heat loads on divertor plates. The FAST divertor will be made of bulk W tiles, for basic operations, but also fully toroidal divertor targets made of liquid lithium (L-Li) are foreseen. Viability tests of such a solution for DEMO divertor will be carried out as final step of an extended program started on FTU tokamak by using a liquid lithium limITER. To have reliable predictions of the thermal loads on the divertor plates and of the core plasma purity a number of numerical self-consistent simulations have been made for the H-mode and steady-state scenario by using the code COREDIV. This code, already validated in the past on experimental data (namely JET, FTU, Textor), is able to describe self-consistently the core and edge plasma in a tokamak device by imposing the continuity of energy and particle fluxes and of particle densities and temperatures at the separatrix. In the present work the results of such calculations will be illustrated, including heat loads on the divertor. The overall picture shows that at the low plasma densities typical of steady state regimes W is effective in dissipating input power by radiative losses, while Li needs additional impurities (Ar, Ne). In the intermediate and, mainly, in the high density H-mode scenarios impurity seeding is needed with either Li or W as target material, but a small (0.08% atomic concentration) amount of Ar, not affecting the core purity, is sufficient to maintain the divertor peak loads below 18 MW/m2 that represents the safety limit for the W monoblock technology, presently accepted for the ITER divertor tiles. The

  6. A modularized operator interface framework for Tokamak based on MVC design pattern

    International Nuclear Information System (INIS)

    Highlights: • Our framework is based on MVC design pattern. • XML is used to cope with minor difference between different applications. • Functions dealing with EPICS and MDSplus have been modularized into reusable modules. • The modularized framework will shorten J-TEXT's software development cycle. - Abstract: Facing various and continually changing experimental needs, the J-TEXT Tokamak experiment requires home-made software applications developed for different sub-systems. Though dealing with different specific problems, these software applications usually share a lot of functionalities in common. With the goal of improving the productivity of research groups, J-TEXT has designed a C# desktop application framework which is mainly focused on operator interface development. Following the Model–View–Controller (MVC) design pattern, the main functionality dealing with Experimental Physics and Industrial Control System (EPICS) or MDSplus has been modularized into reusable modules. Minor difference among applications can be coped with XML configuration files. In this case, developers are able to implement various kinds of operator interface without knowing the implementation details of the bottom functions in Models, mainly focusing on Views and Controllers. This paper presents J-TEXT C# desktop application framework, introducing the technology of fast development of the modularized operator interface. Some experimental applications designed in this framework have been already deployed in J-TEXT, and will be introduced in this paper

  7. 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.

  8. The Physics Basis For An Advanced Physics And Advanced Technology Tokamak Power Plant Configuration, ARIES-ACT1

    Energy Technology Data Exchange (ETDEWEB)

    Charles Kessel, et al

    2014-03-05

    The advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2 and triangularity of 0.63. The broadest pressure cases reached wall stabilized βN ~ 5.75, limited by n=3 external kink mode requiring a conducting shell at b/a = 0.3, and requiring plasma rotation, feedback, and or kinetic stabilization. The medium pressure peaking case reached βN = 5.28 with BT = 6.75, while the peaked pressure case reaches βN < 5.15. Fast particle MHD stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling show that about 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while over 95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring about ~ 1.1 MA of external current drive. This current is supplied with 5 MW of ICRF/FW and 40 MW of LHCD. EC was examined and is most effective for safety factor control over ρ ~ 0.2-0.6 with 20 MW. The pedestal density is ~ 0.9x1020 /m3 and the temperature is ~ 4.4 keV. The H98 factor is 1.65, n/nGr = 1.0, and the net power to LH threshold power is 2.8- 3.0 in the flattop.

  9. Tokamak plasma power balance calculation code (TPC code) outline and operation manual

    International Nuclear Information System (INIS)

    This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)

  10. Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

    Energy Technology Data Exchange (ETDEWEB)

    S.C. Jardin; C.E. Kessel; T.K. Mau; R.L. Miller; F. Najmabadi; V.S. Chan; M.S. Chu; R. LaHaye; L.L. Lao; T.W. Petrie; P. Politzer; H.E. St. John; P. Snyder; G.M. Staebler; A.D. Turnbull; W.P. West

    2003-10-07

    The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.

  11. Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

    International Nuclear Information System (INIS)

    The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented

  12. Influence of various physics phenomena on fast-wave current drive in advanced tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Batchelor, D.B.; Jaeger, E.F.; Carter, M.D.; Goldfinger, R.C.; Stallings, D.C. [Oak Ridge National Lab., TN (United States)

    1992-12-31

    The need for some type of noninductive current drive in advanced tokamaks has been recognized for some time. In reactor-grade plasmas, as envisioned in the International Thermonuclear Experimental Reactor (ITER), high density and temperature may limit the penetration of lower hybrid (LH) waves to only the outer layers of the plasma. Fast waves in the ion cyclotron range of frequencies (ICRF), however, can easily penetrate to the center of such high-density plasmas. With sufficient directivity in the launched wave spectrum, currents can be driven by combined damping of the fast waves on resonant electrons through electron Landau damping (ELD) and transit-time magnetic pumping (TTMP). Experiments to study the feasibility of fast-wave current drive (FWCD) have only recently begun, but theoretical predictions look promising. In this paper we analyze the influence of the relevant physics phenomena, which are not necessarily independent, on current drive performance. Such phenomena include diffraction and other nongeometrical optics processes, k{sub ||} modification, single-pass absorption, and antenna characteristics, such as poloidal extent and poloidal location. To do this, we apply a two-and-one-half dimensional (2 1/2-D), full-wave code (PICES) for modeling ion cyclotron resonance heating (ICRH) and current drive based on the poloidal mode expansion method and the reduced-order expansion. By 2 1/2-D, we mean that 3-D wave fields are calculated in axisymmetric geometry (2-D solution domain - r, {theta}), while the correct toroidal dependence of the antenna source currents is obtained from a 2-D (r, {phi}) recessed antenna code. The model includes the poloidal and toroidal structure of the antennas, the modification of the k{sub ||} spectrum due to the poloidal magnetic field, and a nonperturbative solution for E{sub ||}. A semianalytical model for current drive, including trapped electron effects, is employed. (author) 10 refs., 4 figs.

  13. Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

    International Nuclear Information System (INIS)

    A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the 'advanced tokamak' physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs

  14. Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)

    1994-11-01

    A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.

  15. First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

    Science.gov (United States)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.; Castaldo, C.; De Angeli, M.; Figini, L.; Galperti, C.; Garavaglia, S.; Granucci, G.; Grosso, G.; Korsholm, S. B.; Lontano, M.; Mellera, V.; Minelli, D.; Moro, A.; Nardone, A.; Nielsen, S. K.; Rasmussen, J.; Simonetto, A.; Stejner, M.; Tartari, U.

    2015-10-01

    Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under certain circumstances, the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. Exploiting the front-steering configuration available with the real-time launcher, the implementation of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system during the very first operations in 2014. The present work has been carried out under an EUROfusion Enabling Research project. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

  16. Simple contour analysis of ignition conditions and plasma operating regimes in tokamaks

    International Nuclear Information System (INIS)

    Contour plots of ignition, auxiliary power requirements, heating and operating windows, optimal path to ignition, ignition margin, etc., are generated analytically in terms of a small number of parameters (aB02/q/sub */, R0/B0, , etc.) for classes of devices with equivalent performance. Numerical studies are carried out to map the physics design space. Considering both the Murakami density limit (approx.B0/R0) and the Troyon beta limit (approx.I/aB0), results from analytic calculations indicate that in a standard tokamak geometry (A approx. 2.5 to 3.5, kappa = b/a approx. 1.6 to 1.7, q/sub psi/ approx. 2.6) devices with aB02/q/sub */ approx. 20 should be ignitable provided confinement does not degrade with heating (ohmic + alpha + auxiliary, etc.) power; however, aB02/q/sub */ approx. 30 (25) may be required for minimal ignition for a typical L- (H-) mode confinement scaling. Increased plasma elongation (kappa approx. 2) may help to reduce these requirements

  17. Plasma Fluctuation Studies in the TCV Tokamak: Modeling of Shaping Effects and Advanced Diagnostic Development

    International Nuclear Information System (INIS)

    One of the most important issues for magnetic-confinement fusion research is the so-called anomalous transport across magnetic field lines, i.e. transport that is in excess of that caused by collisional processes. The need to reduce anomalous transport in order to increase the efficiency of a prospective fusion reactor must be addressed through an investigation of its fundamental underlying causes. This thesis is divided into two distinct components: one experimental and instrumental, and the other theoretical and based on numerical modeling. The experimental part consists of the design and installation of a new diagnostic for core turbulence fluctuations in the TCV tokamak. An extensive conceptual investigation of a number of possible solutions, including Beam Emission Spectroscopy, Reflectometry, Cross Polarization, Collective Scattering and different Imaging techniques, was carried out at first. A number of criteria, such as difficulties in data interpretation, costs, variety of physics issues that could be addressed and expected performance, were used to compare the different techniques for specific application to the TCV tokamak. The expected signal to noise ratio and the required sampling frequency for TCV were estimated on the basis of a large number of linear, local gyrokinetic simulations of plasma fluctuations. This work led to the choice of a Zernike phase contrast imaging system in a tangential launching configuration. The diagnostic was specifically designed to provide information on turbulence features up to now unknown. In particular, it is characterized by an outstanding spatial resolution and by the capability to measure a very broad range of fluctuations, from ion to electron Larmor radius scales, thus covering the major part of the instabilities expected to be at play in TCV. The spectrum accessible covers the wavenumber region from 0.9 cm-1 to 60 cm-1 at 24 radial positions with 3 MHz bandwidth. The diagnostic is an imaging technique and is

  18. Tokamak concept innovations

    International Nuclear Information System (INIS)

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  19. Review of ICRF antenna development and heating experiments up to advanced experiment I, 1989 on the JT-60 tokamak

    International Nuclear Information System (INIS)

    Two main subjects of ion cyclotron range of frequencies (ICRF) heating on JT-60 are described in this paper from development phase of the JT-60 ICRF heating system up to advanced experiment I, 1989. One is antenna design and development for the high power JT-60 ICRF heating system (6 MW for 10 s at a frequency range of 108 - 132 MHz). The other is the experimental investigation of characteristics of second harmonic ICRF heating in a large tokamak. (J.P.N.)

  20. Divertor coil power supply in Aditya Tokamak for improved plasma operation

    International Nuclear Information System (INIS)

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a Tokamak with divertor configuration. This moderate field Tokamak is capable of producing 250 kA of plasma current with 300 ms duration. Two new sets of diverter coils will be added to the system with an objective of producing double null plasmas in Aditya Upgrade Tokamak. Diverter coils, made up of continuously transposed conductor, are low voltage high current carrying poloidal field coils. One set of inner divertor coil has radius of 460 mm containing 6 turns and the other set of 1075 mm radius coil with 1 turn makes the outer divertor coils. The simulated plasma double null equilibrium demands 150 kAT of NI for the inner divertor coils and 10 - 20 kAT of NI for outer divertor coils. To energize the divertor coils with required power, a pulsed DC power supply of 3 MW (100V, 30 kA) has been designed. The designed pulsed DC power supply will be a 3-phase, 12-pulse rectifier based convertor power supply having a duty cycle of 300 ms on-time and 15 minutes off-time. The current rise time in the divertor coils will be ∼ 0.6 MA/sec. Detailed design of the divertor power supply with active controls for real time control of the plasma shape will be discussed in this paper. (author)

  1. Characterisation, modelling and control of advanced scenarios in the european tokamak jet; Caracterisation, modelisation et controle des scenarios avances dans le tokamak europeen jet

    Energy Technology Data Exchange (ETDEWEB)

    Tresset, G

    2002-09-26

    The advanced scenarios, developed for less than ten years with the internal transport barriers and the control of current profile, give rise to a 'new deal' for the tokamak as a future thermonuclear controlled fusion reactor. The Joint European Torus (JET) in United Kingdom is presently the most powerful device in terms of fusion power and it has allowed to acquire a great experience in these improved confinement regimes. The reduction of turbulent transport, considered now as closely linked to the shape of current profile optimised for instance by lower hybrid current drive or the self-generated bootstrap current, can be characterised by a dimensionless criterion. Most of useful information related to the transport barriers are thus available. Large database analysis and real time plasma control are envisaged as attractive applications. The so-called 'S'-shaped transport models exhibit some interesting properties in fair agreement with the experiments, while the non-linear multivariate dependencies of thermal diffusivity can be approximated by a neural network, suggesting a new approach for transport investigation and modelling. Finally, the first experimental demonstrations of real time control of internal transport barriers and current profile have been performed on JET. Sophisticated feedback algorithms have been proposed and are being numerically tested to achieve steady-state and efficient plasmas. (author)

  2. Introduction condition of a tokamak fusion power plant as an advanced technology in world energy scenario

    International Nuclear Information System (INIS)

    The present study reveals the following two introduction conditions of a tokamak fusion power plant in a long term world energy scenario. The first condition is the electric breakeven condition, which is required for the fusion energy to be recognized as a suitable candidate of an alternative energy source in the long term world energy scenario. As for the plasma performance (normalized beta value βN, confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density limit fnGW), the electric breakeven condition requires the simultaneous achievement of 1.2NGWtmax=16 T, thermal efficiency ηe=30%, and current drive power PNBIN∼1.8, HH∼1.0, and fnGW∼0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric breakeven condition. The second condition is the economic breakeven condition, which is required to be selected as an alternative energy source. By using a long term world energy and environment model, the potential of the fusion energy in the long term world energy scenario is being investigated. Under the constraint of 550 ppm CO2 concentration in the atmosphere, a breakeven price for introduction of the fusion energy in the year 2050 is estimated from 65mill/kWh to 135mill/kWh, which is considered as the economic breakeven condition in the present study. Under the conditions of Btmax=16T, ηe=40%, plant availability 60%, and a radial build with/without CS coil, the economic breakeven condition requires βN∼2.5 for 135mill/kWh of higher breakeven price case and βN∼6.0 for 65mill/kWh of lower breakeven price case. Finally, the demonstration of steady state operation with βN∼3.0 in the ITER project leads to the prospect to achieve the upper region of breakeven price in the world energy scenario. (author)

  3. Advanced smartgrids for distribution system operators

    CERN Document Server

    Boillot, Marc

    2014-01-01

    The dynamic of the Energy Transition is engaged in many region of the World. This is a real challenge for electric systems and a paradigm shift for existing distribution networks. With the help of "advanced" smart technologies, the Distribution System Operators will have a central role to integrate massively renewable generation, electric vehicle and demand response programs. Many projects are on-going to develop and assess advanced smart grids solutions, with already some lessons learnt. In the end, the Smart Grid is a mean for Distribution System Operators to ensure the quality and the secu

  4. A fast-time-response extreme ultraviolet spectrometer for measurement of impurity line emissions in the Experimental Advanced Superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Ling; Xu, Zong; Wu, Zhenwei; Zhang, Pengfei; Wu, Chengrui; Gao, Wei; Shen, Junsong; Chen, Yingjie; Liu, Xiang; Wang, Yumin; Gong, Xianzu; Hu, Liqun; Chen, Junlin; Zhang, Xiaodong; Wan, Baonian; Li, Jiangang [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230026, Anhui (China); Morita, Shigeru; Ohishi, Tetsutarou; Goto, Motoshi [National Institute for Fusion Science, Toki 509-5292, Gifu (Japan); Department of Fusion Science, Graduate University for Advanced Studies, Toki 509-5292, Gifu (Japan); Dong, Chunfeng [Southwestern Institute of Physics, Chengdu 610041, Sichuan (China); and others

    2015-12-15

    A flat-field extreme ultraviolet (EUV) spectrometer working in the 20-500 Å wavelength range with fast time response has been newly developed to measure line emissions from highly ionized tungsten in the Experimental Advanced Superconducting Tokamak (EAST) with a tungsten divertor, while the monitoring of light and medium impurities is also an aim in the present development. A flat-field focal plane for spectral image detection is made by a laminar-type varied-line-spacing concave holographic grating with an angle of incidence of 87°. A back-illuminated charge-coupled device (CCD) with a total size of 26.6 × 6.6 mm{sup 2} and pixel numbers of 1024 × 255 (26 × 26 μm{sup 2}/pixel) is used for recording the focal image of spectral lines. An excellent spectral resolution of Δλ{sub 0} = 3-4 pixels, where Δλ{sub 0} is defined as full width at the foot position of a spectral line, is obtained at the 80-400 Å wavelength range after careful adjustment of the grating and CCD positions. The high signal readout rate of the CCD can improve the temporal resolution of time-resolved spectra when the CCD is operated in the full vertical binning mode. It is usually operated at 5 ms per frame. If the vertical size of the CCD is reduced with a narrow slit, the time response becomes faster. The high-time response in the spectral measurement therefore makes possible a variety of spectroscopic studies, e.g., impurity behavior in long pulse discharges with edge-localized mode bursts. An absolute intensity calibration of the EUV spectrometer is also carried out with a technique using the EUV bremsstrahlung continuum at 20-150 Å for quantitative data analysis. Thus, the high-time resolution tungsten spectra have been successfully observed with good spectral resolution using the present EUV spectrometer system. Typical tungsten spectra in the EUV wavelength range observed from EAST discharges are presented with absolute intensity and spectral identification.

  5. The Advanced Stellar Compass, Development and Operations

    DEFF Research Database (Denmark)

    Jørgensen, John Leif; Liebe, Carl Christian

    1996-01-01

    this demand the Advanced Stellar Compass (ASC), a fully autonomous miniature star tracker, was developed. This ASC is capable of both solving the "lost in space" problem and determine the attitude with arcseconds precision. The development, principles of operation and instrument autonomy of the ASC...

  6. Regolith Advanced Surface Systems Operations Robot Excavator

    Science.gov (United States)

    Mueller, Robert P.; Smith, Jonathan D.; Ebert, Thomas; Cox, Rachel; Rahmatian, Laila; Wood, James; Schuler, Jason; Nick, Andrew

    2013-01-01

    The Regolith Advanced Surface Systems Operations Robot (RASSOR) excavator robot is a teleoperated mobility platform with a space regolith excavation capability. This more compact, lightweight design (<50 kg) has counterrotating bucket drums, which results in a net-zero reaction horizontal force due to the self-cancellation of the symmetrical, equal but opposing, digging forces.

  7. Advanced Interval Management (IM) Concepts of Operations

    Science.gov (United States)

    Barmore, Bryan E.; Ahmad, Nash'at N.; Underwood, Matthew C.

    2014-01-01

    This document provides a high-level description of several advanced IM operations that NASA is considering for future research and development. It covers two versions of IM-CSPO and IM with Wake Mitigation. These are preliminary descriptions to support an initial benefits analysis

  8. Evaluation of the operational parameters for NBI-driven fusion in low-gain tokamaks with two-component plasma

    Science.gov (United States)

    Chirkov, A. Yu.

    2015-09-01

    Low gain (Q ~ 1) fusion plasma systems are of interest for concepts of fusion-fission hybrid reactors. Operational regimes of large modern tokamaks are close to Q  ≈  1. Therefore, they can be considered as prototypes of neutron sources for fusion-fission hybrids. Powerful neutral beam injection (NBI) can support the essential population of fast particles compared with the Maxwellial population. In such two-component plasma, fusion reaction rate is higher than for Maxwellian plasma. Increased reaction rate allows the development of relatively small-size and relatively inexpensive neutron sources. Possible operating regimes of the NBI-heated tokamak neutron source are discussed. In a relatively compact device, the predictions of physics of two-component fusion plasma have some volatility that causes taking into account variations of the operational parameters. Consequent parameter ranges are studied. The feasibility of regimes with Q  ≈  1 is shown for the relatively small and low-power system. The effect of NBI fraction in total heating power is analyzed.

  9. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  10. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  11. Electron cyclotron emission diagnostics on KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, S. H. [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Daejeon 305-353 (Korea, Republic of); Lee, K. D.; Kwon, M. [National Fusion Research Institute, 113 Gwahangno, Daejeon 305-333 (Korea, Republic of); Kogi, Y. [Fukuoka Institute of Technology, Higashiku, Fukuoka 811-0295 (Japan); Kawahata, K.; Nagayama, Y. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Mase, A. [KASTEC, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  12. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration. PMID:21033954

  13. The Tokamak Fusion Test Reactor D-T modifications and operations

    International Nuclear Information System (INIS)

    This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

  14. An Assessment of the Penetrations in the First Wall Required for Plasma Measurments for Control of an Advanced Tokamak Plasma Demo

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth M. Young

    2010-02-22

    A Demonstration tokamak (Demo) is an essential next step toward a magnetic-fusion based reactor. One based on advanced-tokamak (AT) plasmas is especially appealing because of its relative compactness. However, it will require many plasma measurements to provide the necessary signals to feed to ancillary systems to protect the device and control the plasma. This note addresses the question of how much intrusion into the blanket system will be required to allow the measurements needed to provide the information required for plasma control. All diagnostics will require, at least, the same shielding designs as planned for ITER, while having the capability to maintain their calibration through very long pulses. Much work is required to define better the measurement needs and the quantity and quality of the measurements that will have to be made, and how they can be integrated into the other tokamak structures.

  15. Optical layout and mechanical structure of polarimeter-interferometer system for Experimental Advanced Superconducting Tokamak.

    Science.gov (United States)

    Zou, Z Y; Liu, H Q; Jie, Y X; Ding, W X; Brower, D L; Wang, Z X; Shen, J S; An, Z H; Yang, Y; Zeng, L; Wei, X C; Li, G S; Zhu, X; Lan, T

    2014-11-01

    A Far-InfaRed (FIR) three-wave POlarimeter-INTerferometer (POINT) system for measurement current density profile and electron density profile is under development for the EAST tokamak. The FIR beams are transmitted from the laser room to the optical tower adjacent to EAST via ∼20 m overmoded dielectric waveguide and then divided into 5 horizontal chords. The optical arrangement was designed using ZEMAX, which provides information on the beam spot size and energy distribution throughout the optical system. ZEMAX calculations used to optimize the optical layout design are combined with the mechanical design from CATIA, providing a 3D visualization of the entire POINT system.

  16. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

  17. Magnetic confinement experiment. I: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM`y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nT{tau}`s {approximately} 2.5x greater than ELM`ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices.

  18. Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

    International Nuclear Information System (INIS)

    It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the Dα amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase

  19. Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research

    Science.gov (United States)

    Lee, W.; Park, H. K.; Lee, D. J.; Nam, Y. U.; Leem, J.; Kim, T. K.

    2016-04-01

    The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm-1. The upper limit corresponds to the normalized wavenumber kθρe of ˜0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed.

  20. Simulations of the L-H transition on experimental advanced superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Weiland, Jan [Department Applied Physics, Chalmers University of Technology and Euratom-VR Association, S41296 Gothenburg (Sweden)

    2014-12-15

    We have simulated the L-H transition on the EAST tokamak [Baonian Wan, EAST and HT-7 Teams, and International Collaborators, “Recent experiments in the EAST and HT-7 superconducting tokamaks,” Nucl. Fusion 49, 104011 (2009)] using a predictive transport code where ion and electron temperatures, electron density, and poloidal and toroidal momenta are simulated self consistently. This is, as far as we know, the first theory based simulation of an L-H transition including the whole radius and not making any assumptions about where the barrier should be formed. Another remarkable feature is that we get H-mode gradients in agreement with the α – α{sub d} diagram of Rogers et al. [Phys. Rev. Lett. 81, 4396 (1998)]. Then, the feedback loop emerging from the simulations means that the L-H power threshold increases with the temperature at the separatrix. This is a main feature of the C-mod experiments [Hubbard et al., Phys. Plasmas 14, 056109 (2007)]. This is also why the power threshold depends on the direction of the grad B drift in the scrape off layer and also why the power threshold increases with the magnetic field. A further significant general H-mode feature is that the density is much flatter in H-mode than in L-mode.

  1. Performance and development of the DIII-D tokamak core

    International Nuclear Information System (INIS)

    The DIII-D tokamak is an upgrade of the Doublet III configuration which has operated since early 1986. This paper presents recent advances in performance using the upper divertor, fabrication development for vanadium components, operation of the helium leak checking in a high deuterium background, and restoration of the damaged Ohmic heating solenoid

  2. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  3. Unified Ideal Stability Limits for Advanced Tokamak and Spherical Torus Plasmas

    International Nuclear Information System (INIS)

    Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal nonrotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants

  4. Recent advances in gyrokinetic full-f particle simulation of medium sized Tokamaks with ELMFIRE

    Energy Technology Data Exchange (ETDEWEB)

    Janhunen, S.J.; Kiviniemi, T.P.; Korpio, T.; Leerink, S.; Nora, M. [Helsinki University of Technology, Euratom-Tekes Association, Espoo (Finland); Heikkinen, J.A. [VTT, Euratom-Tekes Association, Espoo (Finland); Ogando, F. [Helsinki University of Technology, Euratom-Tekes Association, Espoo (Finland); Universidad Nacional de Educacion a Distancia, Madrid (Spain)

    2010-05-15

    Large-scale kinetic simulations of toroidal plasmas based on first principles are called for in studies of transition from low to high confinement mode and internal transport barrier formation in the core plasma. Such processes are best observed and diagnosed in detached plasma conditions in mid-sized tokamaks, so gyrokinetic simulations for these conditions are warranted. A first principles test-particle based kinetic model ELMFIRE[1] has been developed and used in interpretation[1,2] of FT-2 and DIII-D experiments. In this work we summarize progress in Cyclone (DIII-D core) and ASDEX Upgrade pedestal region simulations, and show that in simulations the choice of adiabatic electrons results in quenching of turbulence (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  5. Advances of evolutionary computation methods and operators

    CERN Document Server

    Cuevas, Erik; Oliva Navarro, Diego Alberto

    2016-01-01

    The goal of this book is to present advances that discuss alternative Evolutionary Computation (EC) developments and non-conventional operators which have proved to be effective in the solution of several complex problems. The book has been structured so that each chapter can be read independently from the others. The book contains nine chapters with the following themes: 1) Introduction, 2) the Social Spider Optimization (SSO), 3) the States of Matter Search (SMS), 4) the collective animal behavior (CAB) algorithm, 5) the Allostatic Optimization (AO) method, 6) the Locust Search (LS) algorithm, 7) the Adaptive Population with Reduced Evaluations (APRE) method, 8) the multimodal CAB, 9) the constrained SSO method.

  6. Interaction of CLAM Steel with Plasma in HT-7 Tokamak During High Parameter Operation

    Institute of Scientific and Technical Information of China (English)

    LI Chunjing; HUANG Qunying; FENG Yan; LI Jiangang; KONG Mingguang

    2007-01-01

    A Plasma Surface Interaction(PSI)experiment on China Low Activation Martensitic(CLAM)steel was done to check if CLAM steel could be used as a Plasma Facing Material (PFM).A specimen with a diameter of 45 mm was exposed to 897 shots of deuterium plasmas with a total duration of 712 sec at a minor radius of 30 cm in HT-7 tokamak.During the exposure experiment,no observable influence Was found on plasma performance.After exposure,the surface of the specimen seemed as smooth as before but with some colour change at the margin of the specimen.Even though some micro-damage,such as dense blisters,melting,splashing,depositions,and dust,Was found on local surfaces with Scanning Electron Microscopic(SEM)observation.The reflectivity of the specimen decreased only slightly.All of these shows CLAM steel has good stability and irradiation resistance.With further optimization,it could possibly be used as the first mirror material for plasma diagnostics in tokamaks.

  7. Advanced Autonomous Systems for Space Operations

    Science.gov (United States)

    Gross, A. R.; Smith, B. D.; Muscettola, N.; Barrett, A.; Mjolssness, E.; Clancy, D. J.

    2002-01-01

    New missions of exploration and space operations will require unprecedented levels of autonomy to successfully accomplish their objectives. Inherently high levels of complexity, cost, and communication distances will preclude the degree of human involvement common to current and previous space flight missions. With exponentially increasing capabilities of computer hardware and software, including networks and communication systems, a new balance of work is being developed between humans and machines. This new balance holds the promise of not only meeting the greatly increased space exploration requirements, but simultaneously dramatically reducing the design, development, test, and operating costs. New information technologies, which take advantage of knowledge-based software, model-based reasoning, and high performance computer systems, will enable the development of a new generation of design and development tools, schedulers, and vehicle and system health management capabilities. Such tools will provide a degree of machine intelligence and associated autonomy that has previously been unavailable. These capabilities are critical to the future of advanced space operations, since the science and operational requirements specified by such missions, as well as the budgetary constraints will limit the current practice of monitoring and controlling missions by a standing army of ground-based controllers. System autonomy capabilities have made great strides in recent years, for both ground and space flight applications. Autonomous systems have flown on advanced spacecraft, providing new levels of spacecraft capability and mission safety. Such on-board systems operate by utilizing model-based reasoning that provides the capability to work from high-level mission goals, while deriving the detailed system commands internally, rather than having to have such commands transmitted from Earth. This enables missions of such complexity and communication` distances as are not

  8. RASSOR - Regolith Advanced Surface Systems Operations Robot

    Science.gov (United States)

    Gill, Tracy R.; Mueller, Rob

    2015-01-01

    The Regolith Advanced Surface Systems Operations Robot (RASSOR) is a lightweight excavator for mining in reduced gravity. RASSOR addresses the need for a lightweight (robot that is able to overcome excavation reaction forces while operating in reduced gravity environments such as the moon or Mars. A nominal mission would send RASSOR to the moon to operate for five years delivering regolith feedstock to a separate chemical plant, which extracts oxygen from the regolith using H2 reduction methods. RASSOR would make 35 trips of 20 kg loads every 24 hours. With four RASSORs operating at one time, the mission would achieve 10 tonnes of oxygen per year (8 t for rocket propellant and 2 t for life support). Accessing craters in space environments may be extremely hard and harsh due to volatile resources - survival is challenging. New technologies and methods are required. RASSOR is a product of KSC Swamp Works which establishes rapid, innovative and cost effective exploration mission solutions by leveraging partnerships across NASA, industry and academia.

  9. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  10. Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

    Science.gov (United States)

    Xu, J. C.; Wang, L.; Xu, G. S.; Luo, G. N.; Yao, D. M.; Li, Q.; Cao, L.; Chen, L.; Zhang, W.; Liu, S. C.; Wang, H. Q.; Jia, M. N.; Feng, W.; Deng, G. Z.; Hu, L. Q.; Wan, B. N.; Li, J.; Sun, Y. W.; Guo, H. Y.

    2016-08-01

    In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

  11. First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

    DEFF Research Database (Denmark)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.;

    2015-01-01

    Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin...... in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under...... of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system...

  12. First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

    DEFF Research Database (Denmark)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.;

    2015-01-01

    of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system......Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin...... in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under...

  13. New dual gas puff imaging system with up-down symmetry on experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Shao, L. M.; Zweben, S. J.;

    2012-01-01

    Gas puff imaging (GPI) offers a direct and effective diagnostic to measure the edge turbulence structure and velocity in the edge plasma, which closely relates to edge transport and instability in tokamaks. A dual GPI diagnostic system has been installed on the low field side on experimental...... is used to capture the light emission from the image plane with a speed up to 390 804 frames/s with 64x64 pixels and an exposure time of 2.156 mu s. The spatial resolution of the system is 2 mm at the objective plane. A total amount of 200 Pa.L helium gas is puffed into the plasma edge for each GPI...... viewing region for about 250 ms. The new GPI diagnostic has been applied on EAST for the first time during the recent experimental campaign under various plasma conditions, including ohmic, L-mode, and type-I, and type-III ELMy H-modes. Some of these initial experimental results are also presented. © 2012...

  14. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  15. Magnetic confinement experiment -- 1: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1994-12-31

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization.

  16. Operation of a 20 tesla on-axis tokamak toroidal field magnet

    International Nuclear Information System (INIS)

    The Center for Electromechanics at The University of Texas at Austin (CEM-UT) has designed, built, and is presently testing a 20 T on-axis, single turn, toroidal field (TF) coil. The Ignition Technology Demonstration (ITD) is a 0.06-scale IGNITEX (Texas Fusion Ignition Experiment) TF-coil experiment. The purpose of the ITD program is to demonstrate the operation of a 20 T, single turn, TF coil powered by homopolar generators (HPGs). This program is funded by the Advanced Technology Program and the Texas Atomic Energy Research Foundation. Scaling of the prototype 20 T TF coil was selected to be 0.06 on the basis of the maximum current capability of CEM-UT's 60 MJ HPG power supply, which has a rating of 9 MA at 100 V in a parallel configuration. Stresses and temperatures reached in the scale TF coil are representative of those that would be experienced in a full-scale IGNITEX TF coil with a 1.5 m major radius and a 5 s flat top current profile. The 60 MJ HPG system consists of six, 20 MJ, drum-type HPGs each capable of 1.5 MA at 100 V. Only 25% of the available system energy is used to drive the single turn TF coil to 20 T

  17. Advancing Autonomous Operations for Deep Space Vehicles

    Science.gov (United States)

    Haddock, Angie T.; Stetson, Howard K.

    2014-01-01

    Starting in Jan 2012, the Advanced Exploration Systems (AES) Autonomous Mission Operations (AMO) Project began to investigate the ability to create and execute "single button" crew initiated autonomous activities [1]. NASA Marshall Space Flight Center (MSFC) designed and built a fluid transfer hardware test-bed to use as a sub-system target for the investigations of intelligent procedures that would command and control a fluid transfer test-bed, would perform self-monitoring during fluid transfers, detect anomalies and faults, isolate the fault and recover the procedures function that was being executed, all without operator intervention. In addition to the development of intelligent procedures, the team is also exploring various methods for autonomous activity execution where a planned timeline of activities are executed autonomously and also the initial analysis of crew procedure development. This paper will detail the development of intelligent procedures for the NASA MSFC Autonomous Fluid Transfer System (AFTS) as well as the autonomous plan execution capabilities being investigated. Manned deep space missions, with extreme communication delays with Earth based assets, presents significant challenges for what the on-board procedure content will encompass as well as the planned execution of the procedures.

  18. Operational experience at the Advanced Light Source

    International Nuclear Information System (INIS)

    The Advanced Light Source (ALS) has been operational for users since October 1993 when white light from a bend magnet was delivered to the Center for X-Ray Optic close-quote s (CXRO) x-ray microprobe end station. Since then, the ALS has installed and commissioned three undulators and their beamlines (including monochromators and post-monochromator focusing optics), and eight bend-magnet beamlines, including one dedicated to machine diagnostics. Apart from one serious outage, when scheduled beam was not available to users for 17 days, the ALS has enjoyed remarkable operating statistics, with typically 95% of scheduled beam time delivered to the users. Beam quality has also been very good. With a vertical emittance measured at 0.06 nm-rad, the electron beam is kept stable to about one-tenth of its transverse dimensions, in the face of changing error fields in the insertion devices (as their main fields are varied), temperature variations, and floor vibration. The longitudinal motion of the beam, which leads to an increase in the electron beam energy spread and thence to a degradation of the undulator spectra, has recently been brought under control by the addition of an innovative feedback system. This paper focuses on those aspects of electron beam stability that we find most affect the ALS users: beam size and position, and energy spread. copyright 1996 American Institute of Physics

  19. Operational experiences at the advanced light source

    International Nuclear Information System (INIS)

    The Advanced Light Source (ALS) has been operational for users since October 1993 when white light from a bend magnet was delivered to the Center for X-Ray Optic's (CXRO) x-ray microprobe end-station. Since then, the ALS has installed and commissioned three undulators and their beamlines (including monochrornators and post-monochromator focusing optics), and eight bend magnet beamlines, including one dedicated to machine diagnostics. Apart from one serious outage, when scheduled beam was not available to users for 17 days, the ALS has enjoyed remarkable operating statistics, with typically 95% of scheduled beam-time delivered to the users. Beam quality has also been very good. With a vertical emittance measured at 0.06 nm-rad, the electron beam is kept stable to about one-tenth of it's transverse dimensions, in the face of changing error fields in the insertion devices (as their main fields are varied), temperature variations and floor vibration. The longitudinal motion of the beam, which leads to an increase in the electron beam energy spread, and thence, to a degradation of the undulator spectra, has recently been brought under control by the addition of an innovative feedback system. This paper focuses on those aspects of electron beam stability that we find most affect the ALS users: beam size and position, and energy spread

  20. Development of advanced inductive scenarios for ITER

    NARCIS (Netherlands)

    Luce, T. C.; Challis, C. D.; Ide, S.; Joffrin, E.; Kamada, Y.; Polizer, P. A.; Schweinzer, J.; Sips, A.C.C.; Stober, J.; Giruzzi, G.; Kessel, C. E.; Murakami, M.; Na, Y.-S.; Park, J. M.; Polevoi, A. R.; Budny, R. V.; Citrin, J.; Garcia, J.; Hayashi, N.; Hobirk, J.; Hudson, B. F.; Imbeaux, F.; Isayama, A.; McDonald, D. C.; Nakano, T.; Oyama, N.; Parail, V.V.; Petrie, T. W.; Petty, C. C.; Suzuki, T.; Wade, M. R.

    2014-01-01

    Since its inception in 2002, the International Tokamak Physics Activity topical group on Integrated Operational Scenarios (IOS) has coordinated experimental and modelling activity on the development of advanced inductive scenarios for applications in the ITER tokamak. The physics basis and the prosp

  1. High-beta operation and magnetohydrodynamic activity on the TFTR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, K.; Arunasalam, V.; Barnes, C. W.; Bell, M. G.; Bitter, M.; Boivin, R.; Bretz, N. L.; Budny, R.; Bush, C. E.; Cavallo, A.; Chu, T. K.; Cohen, S. A.; Colestock, P.; Davis, S. L.; Dimock, D. L.; Dylla, H. F.; Efthimion, P. C.; Ehrhrardt, A. B.; Fonck, R. J.; Fredrickson, E.; Furth, H. P.; Gammel, G.; Goldston, R. J.; Greene, G.; Grek, B.; Grisham, L. R.; Hammett, G.; Hawryluk, R. J.; Hendel, H. W.; Hill, K. W.; Hinnov, E.; Hoffman, D. J.; Hosea, J.; Howell, R. B.; Hsuan, H.; Hulse, R. A.; Janos, A. C.; Jassby, D.; Jobes, F.; Johnson, D. W.; Johnson, L. C.; Kaita, R.; Kieras-Phillips, C.; Kilpatrick, S. J.; LaMarche, P. H.; LeBlanc, B.; Manos, D. M.; Mansfield, D. K.; Mazzucato, E.; McCarthy, M. P.; McCune, M. C.; McNeill, D. H.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Monticello, D.; Motley, R.; Mueller, D.; Murphy, J. A.; Nagayama, Y.; Nazakian, D. R.; Neischmidt, E. B.; Owens, D. K.; Park, H.; Park, W.; Pitcher, S.; Ramsey, A. T.; Redi, M. H.; Roquemore, A. L.; Rutherford, P. H.; Schilling, G.; Schivell, J.; Schmidt, G. L.; Scott, S. D.; Sinnis, J. C.; Stevens, J.; Stratton, B. C.; Stodiek, W.; Synakowski, E. J.; Tang, W. M.; Taylor, G.; Timberlake, J. R.; Towner, H. H.; Ulrickson, M.; von Goeler, S.; Wieland, R.; Williams, M.; Wilson, J. R.; Wong, K.-L.; Yamada, M.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zweben, S. J.

    1990-01-01

    Magnetohydrodynamic (MHD) activity within three zones (core, half-radius, and edge) of TFTR ( Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 51) tokamak plasmas are discussed. Near the core of the plasma column, sawteeth are often observed. Two types of sawteeth are studied in detail; one with complete, and the other with incomplete, magnetic reconnection. Their characteristics are determined by the shape of the q profile. Near the half-radius the m/n =3/2 and 2/1 resistive ballooning modes are found to correlate with a beta collapse. The pressure and the pressure gradient at the mode rational surface are found to play an important role in stability. MHD activity is also studied at the plasma edge during limiter H modes. The edge localized modes (ELM's) are found to have a precursor mode with a frequency between 50--500 kHz and a mode number m/n=1/0. The mode does not show a ballooning structure. While these instabilities have been studied on many other machines, on TFTR the studies have been extended to high pressure (plasma pressure greater than 4 x 10⁵ Pa) and low collisionality ( vi*(a/2) < 0.002, ve* ( a/2)< 0.01).

  2. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    Science.gov (United States)

    Sager, G. T.; Wong, C. P. C.; Kapich, D. D.; McDonald, C. F.; Schleicher, R. W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.

  3. Design and optimization of Artificial Neural Networks for the modelling of superconducting magnets operation in tokamak fusion reactors

    Science.gov (United States)

    Froio, A.; Bonifetto, R.; Carli, S.; Quartararo, A.; Savoldi, L.; Zanino, R.

    2016-09-01

    In superconducting tokamaks, the cryoplant provides the helium needed to cool different clients, among which by far the most important one is the superconducting magnet system. The evaluation of the transient heat load from the magnets to the cryoplant is fundamental for the design of the latter and the assessment of suitable strategies to smooth the heat load pulses, induced by the intrinsically pulsed plasma scenarios characteristic of today's tokamaks, is crucial for both suitable sizing and stable operation of the cryoplant. For that evaluation, accurate but expensive system-level models, as implemented in e.g. the validated state-of-the-art 4C code, were developed in the past, including both the magnets and the respective external cryogenic cooling circuits. Here we show how these models can be successfully substituted with cheaper ones, where the magnets are described by suitably trained Artificial Neural Networks (ANNs) for the evaluation of the heat load to the cryoplant. First, two simplified thermal-hydraulic models for an ITER Toroidal Field (TF) magnet and for the ITER Central Solenoid (CS) are developed, based on ANNs, and a detailed analysis of the chosen networks' topology and parameters is presented and discussed. The ANNs are then inserted into the 4C model of the ITER TF and CS cooling circuits, which also includes active controls to achieve a smoothing of the variation of the heat load to the cryoplant. The training of the ANNs is achieved using the results of full 4C simulations (including detailed models of the magnets) for conventional sigmoid-like waveforms of the drivers and the predictive capabilities of the ANN-based models in the case of actual ITER operating scenarios are demonstrated by comparison with the results of full 4C runs, both with and without active smoothing, in terms of both accuracy and computational time. Exploiting the low computational effort requested by the ANN-based models, a demonstrative optimization study has been

  4. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  5. Real-Time Control of Tokamak Plasmas: from Control of Physics to Physics-Based Control

    OpenAIRE

    Felici, Federico

    2011-01-01

    Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solut...

  6. The EuroGEOSS Advanced Operating Capacity

    Science.gov (United States)

    Nativi, S.; Vaccari, L.; Stock, K.; Diaz, L.; Santoro, M.

    2012-04-01

    The concept of multidisciplinary interoperability for managing societal issues is a major challenge presently faced by the Earth and Space Science Informatics community. With this in mind, EuroGEOSS project was launched on May 1st 2009 for a three year period aiming to demonstrate the added value to the scientific community and society of providing existing earth observing systems and applications in an interoperable manner and used within the GEOSS and INSPIRE frameworks. In the first period, the project built an Initial Operating Capability (IOC) in the three strategic areas of Drought, Forestry and Biodiversity; this was then enhanced into an Advanced Operating Capacity (AOC) for multidisciplinary interoperability. Finally, the project extended the infrastructure to other scientific domains (geology, hydrology, etc.). The EuroGEOSS multidisciplinary AOC is based on the Brokering Approach. This approach aims to achieve multidisciplinary interoperability by developing an extended SOA (Service Oriented Architecture) where a new type of "expert" components is introduced: the Broker. These implement all mediation and distribution functionalities needed to interconnect the distributed and heterogeneous resources characterizing a System of Systems (SoS) environment. The EuroGEOSS AOC is comprised of the following components: • EuroGEOSS Discovery Broker: providing harmonized discovery functionalities by mediating and distributing user queries against tens of heterogeneous services; • EuroGEOSS Access Broker: enabling users to seamlessly access and use heterogeneous remote resources via a unique and standard service; • EuroGEOSS Web 2.0 Broker: enhancing the capabilities of the Discovery Broker with queries towards the new Web 2.0 services; • EuroGEOSS Semantic Discovery Broker: enhancing the capabilities of the Discovery Broker with semantic query-expansion; • EuroGEOSS Natural Language Search Component: providing users with the possibilities to search for

  7. Advances in tokamak control: from multi-actuator MHD control to model-based current profile tailoring

    Science.gov (United States)

    Felici, Federico

    2012-10-01

    Recent experiments on TCV have demonstrated integrated control of the sawtooth and Neoclassical Tearing Mode (NTM) instabilities in a combined preemption-suppression strategy. This strategy is enabled by new sawtooth control methods (sawtooth pacing) in which modulation of sawtooth-stabilizing electron cyclotron power during the sawtooth cycle stimulates the advent of the crash. Rather than controlling the average sawtooth period, the precise timing of each individual crash can now be prescribed. Using this knowledge, efficient preemptive stabilization of NTMs becomes possible by applying power on the rational surface only at the instant of the crash-generating seed island. TCV experiments demonstrate that this approach, reinforced by NTM stabilization as a backup strategy, is effectively failsafe. This opens the road to inductive H-mode scenarios with long sawteeth providing longer inter-crash periods of high density and temperature. Also Edge Localized Modes are susceptible to EC modulation and it is shown that individual ELM events can be controlled using similar techniques. For advanced tokamak scenarios, MHD control is to be combined with optimization and control of the plasma kinetic and magnetic profile evolution in time. Real-time simulation of a physical model (RAPTOR) of current transport, including bootstrap current, neoclassical conductivity and auxiliary current drive, yields complete knowledge of the relevant profiles at any given time. The pilot implementation on TCV shows that these calculations can indeed be done in real-time and the resulting profiles have been included in feedback control schemes. Integration of this model with time-varying equilibria and internal current profile diagnostics provides a new framework for real-time interpretation of diagnostic data for plasma prediction, scenario monitoring, disruption prevention and feedback control.

  8. Radioactivity evaluation for the KSTAR tokamak

    International Nuclear Information System (INIS)

    The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 1016 s-1 through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10-4 mrem h-1 in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 106 Bq kg-1 in the carbon graphite of a plasma-facing wall. (authors)

  9. Design and Analysis of the Thermal Shield of EAST Tokamak

    Science.gov (United States)

    Xie, Han; Liao, Ziying

    2008-04-01

    EAST (Experimental Advanced Superconducting Tokamak) is a tokamak with superconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.

  10. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  11. Experiment and Operation of a LHCD-35 kV/2.8 MW/1000 s High-Voltage Power Supply on HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    黄懿赟

    2002-01-01

    A -35 kV/2.8 MW/1000s high-voltage power supply (HVPS) for HT-7 superconducting tokamak has been built successfully. The HVPS is scheduled to run on a 2.45 GHz/1 MW lower hybrid current drive (LHCD) [1] system of HT-7 superconducting tokamak before the set-up of HT-7 superconducting tokamak in 2003. The HVPS has a series of advantages such as good steady and dynamic response, logical computer program controlling the HVPS without any fault, operational panel and experimental board for data acquisition, which both are grounded distinctively in a normative way to protect the main body of HVPS along with its attached equipments from dangers. Electric power cables and other control cables are disposed reasonably, to prevent signals from magnetic interference and ensure the precision of.signal transfer.This paper involves the experiment and operation of a 35 kV/2.8 MW/1000 s HVPS [2] for 2.45 GHz/1 MW LHCD system. The reliability and feasibility of the HVPS has been demonstrated in comparison with experimental results of original design and simulation data.

  12. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  13. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  14. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  15. DIII-D research operations

    International Nuclear Information System (INIS)

    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

  16. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  17. Preliminary Study of Ideal Operational MHD Beta Limit in HL-2A Tokamak Plasmas

    Institute of Scientific and Technical Information of China (English)

    SHEN Yong; DONG Jiaqi; HE Hongda; A. D. TURNBULL

    2009-01-01

    Magnetohydrodynamic (MHD) n=1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of βcN=2.56 for a typical HL-2A discharge with a plasma current Ip=0.245 MA, and the scaling of βcN~constant is confirmed.

  18. Initial operation of ECRH heating experiments on the Versator II tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Luckhardt, S.C.; Chen, K.I.; Kirkwood, R.; Porkolab, M.; Singleton, D.; Squire, J.; Villasenor, J.; Lu, Z.

    1987-09-01

    Operation of a 35GHz electron cyclotron heating experiment has begun on Versator II with gyrotron power of 100kW. The EC antenna is located on the high magnetic field side of the plasma and launches linearly polarized radiation in the HE11 hybrid mode with externally controllable polarization and parallel index of refraction. The transmission system provides mode conversion from the TE01 output mode of the gyrotron to the HE11 mode and polarization control. The mode transformation characteristics of the transmission system were measured by means of a computer controlled two dimensional scanning system, and contour plots of the far field radiation pattern of each transmission system element were made and compared with theory. Overall the transmission system is found to be approximately 95% efficient with mode patterns in generally excellent agreement with theory.

  19. Preliminary Study of Ideal Operational MHD Beta Limit in HL-2A Tokamak Plasmas

    Science.gov (United States)

    Shen, Yong; Dong, Jiaqi; He, Hongda; D. Turnbull, A.

    2009-04-01

    Magnetohydrodynamic (MHD) n = 1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of βcN = 2.56 for a typical HL-2A discharge with a plasma current Ip = 0.245 MA, and the scaling of βcN ~constant is confirmed.

  20. Spectroscopic observation of temperature and density modulations in the boundary layer during ergodic divertor operation in the Tore Supra tokamak

    International Nuclear Information System (INIS)

    Spatially resolved spectroscopic measurements of carbon impurity ion line brightness profiles, in ergodic divertor (ED) Tore Supra tokamak plasmas, have shown poloidal electron temperature and density modulations in the peripheral ED layer. These effects have been qualitatively reproduced by using a field line tracing code. (author). 14 refs., 3 figs

  1. Major progress on tore supra toward steady state operation of tokamaks

    International Nuclear Information System (INIS)

    During winter 2000-2001, a major upgrade of the internal components of Tore Supra has been completed that increased the heat extraction capability to 25 MW in steady state. Operating Tore Supra in this new configuration has produced a wealth of new results. The highlights of the 2002 long duration discharges campaign are: 4 minutes 25 seconds long discharges with an integrated energy of 0.75 GJ, which is three time higher than the old Tore Supra world record; recharge of the primary transformer by Lower Hybrid Current Drive (LHCD) for about 1 minute; 4 minutes long LHCD pulses; 1 minute long Ion Cyclotron Resonant Heating (ICRH) pulse (0.11 GJ of ICRH injected energy). Beyond the quantitative step, significant qualitative progress in the steady state nature of the discharge has been accomplished: contrary to the situation in the old Tore Supra configuration, the plasma density is perfectly controlled by active pumping over the overall shot duration. The duration of Tore Supra discharges is sufficient to allow the complete diffusion of the resistive current. Surprising new physics is revealed in such discharges when approaching zero loop voltage. Slow central electron temperature oscillations have been observed in a variety of situations. Such oscillations are not likely to be linked to any MHD instabilities and probably results from an interplay between current profile shape, LHCD power deposition and transport. Analysis of the temperature gradient in the core region shows a very interesting behaviour and the normalised temperature gradient length is compared to the critical thresholds. Finally, the performance of heating and current drive systems and the observations made of the interior of Tore Supra after the long duration discharges campaign are reported. (author)

  2. Status and prospect of JT-60 plasma control and diagnostic data processing systems for advanced operation scenarios

    International Nuclear Information System (INIS)

    The large tokamak fusion device JT-60 is expected to explore more advanced tokamak discharge scenarios toward the ITER and a future power reactor. Since various experimental issues are to be adequately discussed, and possibly to be solved in JT-60, the plasma real-time control system has been drastically improved with remodelling in hardware as well as in software. To satisfy the requirements, a 'multiple networks' structure is employed as a basic principle of the remodelling. Distributed processors for diagnostics, actuators, and supervisory controllers are linked through a reflective memory (RM) network for fast, real-time communication. Similarly, advanced and complex calculations to reproduce plasma shape and profiles are performed by several processors connected to the same RM network. In this report, we discuss the developments to improve the JT-60 plasma control and data processing systems. In addition, a future plasma control system leading to a standard design for a power reactor is outlined, which is based on the 20-year plasma operation experience

  3. Transport of Dust Particles in Tokamak Devices

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A Y; Smirnov, R D; Krasheninnikov, S I; Rognlien, T D; Rozenberg, M

    2006-06-06

    Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.

  4. Advances in the giga-joule operation of Tore Supra

    International Nuclear Information System (INIS)

    Full text of publication follows. Integrating all technological elements required for long pulse operation (superconducting magnets, actively cooled plasma facing components, long pulse heating and current drive systems), the Tore Supra tokamak routinely addresses the physics and technology issues related to the steady state operation of magnetic fusion devices. During the last experimental campaign, the recently upgraded heating and current drive system has been extensively used to explore steady-state scenarios in an expanded operational range. The lower hybrid (LH), ion cyclotron (IC) and electron cyclotron (EC) systems have been successfully combined in stationary shots (duration ∼ 150 s, injected power up to ∼ 8 MW, injected/extracted energy up to ∼ 1 GJ). Injection of LH power in the 5.0-5.7 MW range has extended the domain of accessible plasma parameters to ∼ 3.0*1019 m-3, Ip ∼ 0.7 MA (βp ∼ 0.6, βN ∼ 0.7) with high non-inductive current fraction (∼80%). These discharges also exhibit steady electron internal transport barriers (ITB). We report on various issues relevant to steady-state operation on fusion reactors, ranging from operational aspects and limitations related to the achievement of long pulses in a fully actively cooled fusion device (e.g. overheating due to fast particle losses, real-time protection of plasma facing components, route to MHD stability...), to more fundamental plasma physics topics. The latter includes a beneficial influence of IC heating on the MHD stability in these discharges, which has been studied in details. Another interesting observation is the appearance of oscillations of the central temperature with typical periods of the order of one to several seconds, caused by a non-linear interplay between LH deposition, MHD activity and bootstrap current in the presence of an ITB. (authors)

  5. SPECIAL TOPIC: A two-time-scale dynamic-model approach for magnetic and kinetic profile control in advanced tokamak scenarios on JET

    Science.gov (United States)

    Moreau, D.; Mazon, D.; Ariola, M.; DeTommasi, G.; Laborde, L.; Piccolo, F.; Sartori, F.; Tala, T.; Zabeo, L.; Boboc, A.; Bouvier, E.; Brix, M.; Brzozowski, J.; Challis, C. D.; Cocilovo, V.; Cordoliani, V.; Crisanti, F.; DeLa Luna, E.; Felton, R.; Hawkes, N.; King, R.; Litaudon, X.; Loarer, T.; Mailloux, J.; Mayoral, M.; Nunes, I.; Surrey, E.; Zimmerman, O.; EFDA Contributors, JET

    2008-10-01

    Real-time simultaneous control of several radially distributed magnetic and kinetic plasma parameters is being investigated on JET, in view of developing integrated control of advanced tokamak scenarios. This paper describes the new model-based profile controller which has been implemented during the 2006-2007 experimental campaigns. The controller aims to use the combination of heating and current drive (H&CD) systems—and optionally the poloidal field (PF) system—in an optimal way to regulate the evolution of plasma parameter profiles such as the safety factor, q(x), and gyro-normalized temperature gradient, \\rho _Te^*(x) . In the first part of the paper, a technique for the experimental identification of a minimal dynamic plasma model is described, taking into account the physical structure and couplings of the transport equations, but making no quantitative assumptions on the transport coefficients or on their dependences. To cope with the high dimensionality of the state space and the large ratio between the time scales involved, the model identification procedure and the controller design both make use of the theory of singularly perturbed systems by means of a two-time-scale approximation. The second part of the paper provides the theoretical basis for the controller design. The profile controller is articulated around two composite feedback loops operating on the magnetic and kinetic time scales, respectively, and supplemented by a feedforward compensation of density variations. For any chosen set of target profiles, the closest self-consistent state achievable with the available actuators is uniquely defined. It is reached, with no steady state offset, through a near-optimal proportional-integral control algorithm. Conventional optimal control is recovered in the limiting case where the ratio of the plasma confinement time to the resistive diffusion time tends to zero. Closed-loop simulations of the controller response have been performed in

  6. Regolith Advanced Surface Systems Operations Robot (RASSOR)

    Science.gov (United States)

    Mueller, Robert P.; Smith, Jonathan D.; Cox, Rachel E.; Schuler, Jason M.; Ebert, Tom; Nick, Andrew J.

    2012-01-01

    Regolith is abundant on extra-terrestrial surfaces and is the source of many resources such as oxygen, hydrogen, titanium, aluminum, iron, silica and other valuable materials, which can be used to make rocket propellant, consumables for life support, radiation protection barrier shields, landing pads, blast protection berms, roads, habitats and other structures and devices. Recent data from the Moon also indicates that there are substantial deposits of water ice in permanently shadowed crater regions and possibly under an over burden of regolith. The key to being able to use this regolith and acquire the resources, is being able to manipulate it with robotic excavation and hauling machinery that can survive and operate in these very extreme extra-terrestrial surface environments. In addition, the reduced gravity on the Moon, Mars, comets and asteroids poses a significant challenge in that the necessary reaction force for digging cannot be provided by the robot's weight as is typically done on Earth. Space transportation is expensive and limited in capacity, so small, lightweight payloads are desirable, which means large traditional excavation machines are not a viable option. A novel, compact and lightweight excavation robot prototype for manipulating, excavating, acquiring, hauling and dumping regolith on extra-terrestrial surfaces has been developed and tested. Lessons learned and test results will be presented including digging in a variety of lunar regolith simulant conditions including frozen regolith mixed with water ice.

  7. Advanced solutions for operational reliability improvements

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, K. [VTT Manufacturing Technology, Espoo (Finland)

    1997-12-31

    A great number of new technical tools are today developed for improved operational reliability of machines and industrial equipment. Examples of such techniques and tools recently developed at the Technical Research Centre of Finland (VTT) are: metallographic approach for steam-piping lifetime estimation, an expert system AURORA for corrosion prediction and material selection, an automatic image-processing-based on-line wear particle analysis system, microsensors for condition monitoring, a condition monitoring and expert system, CEPDIA, for the diagnosis of centrifugal pumps, a machine tool analysis and diagnostic expert system, non-leakage magnetic fluid seals with extended lifetime and diamond-like surface coatings on components with decreased friction and wear properties. A hyperbook-supported holistic approach to problem solving in maintenance and reliability engineering has been developed to help the user achieve a holistic understanding of the problem and its relationships, to navigate among the several technical tools and methods available, and to find those suitable for his application. (orig.)

  8. Two phase liquid helium flow testing to simulate the operation of a cryocondensation pump in the D3-D tokamak

    Science.gov (United States)

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

    1994-06-01

    A liquid helium-cooled cryocondensation pump has been installed in the D3-D tokamak fusion energy research experiment at General Atomics. The pump is located within the tokamak vacuum chamber beneath the divertor baffle plates and is utilized for plasma density and contamination control. Two-phase helium flows through the pump at 5 to 10 g/s utilizing the heat transfer and constant temperature characteristics of boiling liquid . helium. The pump is designed for a pumping speed of 32,000 1/s. Extensive testing was performed with a prototypical pump test fixture. Several pump geometries (simple tube, coaxial flow plug, and coaxial slotted insert) were tested, in an iterative process, to determine which was the most satisfactory for stable cryocondensation pumping. Results from the different tests illustrating the temperature distribution and flow characteristics for each configuration are presented.

  9. Two phase liquid helium flow testing to simulate the operation of a cryocondensation pump in the DIII-D tokamak

    Science.gov (United States)

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

    A liquid helium-cooled cryocondensation pump has been installed in the DIII=D tokamak fusion energy research experiment at General Atomics. The pump is located within the tokamak vacuum chamber beneath the divertor baffle plates and is utilized for plasma density and contamination control. Two-phase helium flows through the pump at 5 to 10 g/s utilizing the beat transfer and constant temperature characteristics of boiling liquid helium. The pump is designed for a pumping speed of 32,0001/s. Extensive testing was performed with a prototypical pump test fixture. Several pump geometries (simple tube, coaxial flow plug, and coaxial slotted insert) were tested, in an iterative process, to determine which was the most satisfactory for stable cryocondensation pumping. Results from the different tests illustrating the temperature distribution and flow characteristics for each configuration are presented.

  10. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cTe/eB(δni/ni)rms which is also derived by a simple theory, the cross-field diffusion time, tp=a2/D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  11. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  12. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

    2003-12-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  13. Elements of an advanced integrated operator control station

    International Nuclear Information System (INIS)

    One of the critical determinants of peformance for any remotely operated maintenance system is the compatibility achieved between elements of the man/machine interface (e.g., master manipulator controller, controls, displays, etc.) and the human operator. In the Remote Control Engineering task of the Consolidated Fuel Reprocessing Program, considerable attention has been devoted to optimizing the man/machine interface of the operator control station. This system must be considered an integral element of the overall maintenance work system which includes transporters, manipulators, remote viewing, and other parts. The control station must reflect the integration of the operator team, control/display panels, manipulator master controllers, and remote viewing monitors. Human factors principles and experimentation have been used in the development of an advanced integrated operator control station designed for the advance servomanipulator. Key features of this next-generation design are summarized in this presentation. 7 references, 4 figures

  14. The Advanced Photon Source: Performance and results from early operation

    Energy Technology Data Exchange (ETDEWEB)

    Moncton, D.E. [Argonne National Lab., IL (United States). Advanced Photon Source

    1997-10-01

    The Advanced Photon Source at Argonne National Laboratory is now providing researchers with extreme-brilliance undulator radiation in the hard x-ray region of the spectrum. All technical facilities and components are operational and have met design specifications. Fourteen research teams, occupying 20 sectors on the APS experiment hall floor, are currently installing beamline instrumentation or actively taking data. An overview is presented for the first operational years of the Advanced Photon Source. Emphasis is on the performance of accelerators and insertion devices, as well as early scientific results and future plans.

  15. Tritium catalyzed deuterium tokamaks

    International Nuclear Information System (INIS)

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the 3He from the D(D,n)3He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general)

  16. Challenges to radiative divertor/mantle operations in advanced, steady-state scenarios

    International Nuclear Information System (INIS)

    Full text of publication follows. Managing the heat exhaust problem is well recognized to be a major challenge in transforming present successes in magnetic confinement fusion experiments to demonstration of cost-effective, steady-state power generation from fusion [1][2]. One approach is to convert plasma thermal energy, normally directed to isolated surfaces, to isotropic photon emission, distributing exhaust power over a large surface area. Successful demonstrations of this technique on existing short pulse devices are shown, along with the inherent limitations; the collapse of core confinement with excessive radiation from the bulk plasma and restrictions to dissipation in the divertor volume. Feedback control of impurity seeding is discussed, showing recent examples from tokamaks [3]. For steady-state devices, additional constraints on divertor scenarios are driven by long-term plasma material interaction effects, with fuel recycling, net erosion limits and surface morphology changes forcing detached plasma operation where both heat and particle fluxes are substantially reduced. The instability of these detachment layers in standard X-point divertors with impurity seeding is outlined. Achieving these steady-state, high performance scenarios also restricts the divertor solution by requiring it be compatible with current-drive actuators and enhanced core confinement regimes. While ITER will operate with impurity seeding in a conventional tokamak geometry [4], it is not clear that this concept will reliably scale to a reactor and has been identified as a major risk factor in the development of fusion power [2]. Alternatives concepts are discussed, including the snowflake [5] and super-X divertor [6], along with their respective proof of principle experiments. The complications in convincingly scaling these concepts to a reactor are outlined, including challenges in validating numerical simulations of advanced, dissipative divertors. References: [1] Greenwald, M

  17. Operating Experiences with an Advanced Fabric Energy Storage System

    Directory of Open Access Journals (Sweden)

    R.J Fuller

    2012-11-01

    Full Text Available Despite their proven track record in the cold climate countries of northern Europe, there are no reports in the research literature of experiences using advanced fabric energy storage (FES systems in countries where cooling rather than heating is the main priority. This paper reports some of the experiences with the first known advanced FES system in Australia made over the first full calendar year of operation. It is located in a three-storey building on a university campus in Victoria and has been in operation since mid-2002. Temperature, energy use and operational mode data were recorded during 2003. Airflow measurements through the FES system have been made in five areas of the building. On-going operating problems still exist with the system and this has prevented a conclusive evaluation of its suitability for the southern Australian climate.

  18. Industry roles in the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    There are several distinguishing features of the Tokamak Physics Experiment (TPX) to be found in the TPX program and in the organizations for constructing and operating the machine. Programmatically, TPX addresses several issues critical to the viability of magnetic fusion power plants. Organizationally, it is a multi-institutional partnership to construct and operate the machine and carry out its program mission. An important part of the construction partnership is the integrated industrial responsibility for design, R ampersand D, and construction. The TPX physics design takes advantage of recent research on advanced tokamak operating modes achieved for time scales of the order of seconds that are consistent with continuous operation. This synergism of high performance (higher power density) modes with plasma current driven mostly by internal pressure (boot-strap effect) points toward tokamak power plants that will be cost-competitive and operate continuously. A large fraction of the project is subcontracted to industry. By policy, these contracts are at a high level in the project breakdown of work, giving contractors much of the overall responsibility for a given major system. That responsibility often includes design and R ampersand D in addition to the fabrication of the system in question. Each contract is managed through one of three national laboratories: PPPL, LLNL, and ORNL. Separate contracts for system integration and construction management round out the industry involvement in the project. This integrated, major responsibility attracts high-level corporate attention within each company, which are major corporations with long-standing interest in fusion. Through the contracts already established on the TPX project, a new standard for industry involvement in fusion has been set, and these industries will be well prepared for future fusion projects

  19. Advanced measurement approach with loss distribution in operational risk management

    OpenAIRE

    Atilla ÇİFTER; Chambers, Nurgül

    2007-01-01

    According to the last proposal by Basel Committee, commercial banks are allowed to use advanced measurement approach for operational risk. Since basic indicator and standard approach considers operational risk as a percentage of gross profit, these methodologies are not satisfactory as real lost or probability of lost are not taken into consideration. In this article, loss distribution approach is applied with simulated data. 20 nonparametric loss distributions and mixing internal and externa...

  20. Fabrication and Characterization of Samples for a Material Migration Experiment on the Experimental Advanced Superconducting Tokamak (EAST).

    Energy Technology Data Exchange (ETDEWEB)

    Wampler, William R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Van Deusen, Stuart B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    This report documents work done for the ITER International Fusion Energy Organization (Sponsor) under a Funds-In Agreement FI 011140916 with Sandia National Laboratories. The work consists of preparing and analyzing samples for an experiment to measure material erosion and deposition in the EAST Tokamak. Sample preparation consisted of depositing thin films of carbon and aluminum onto molybdenum tiles. Analysis consists of measuring the thickness of films before and after exposure to helium plasma in EAST. From these measurements the net erosion and deposition of material will be quantified. Film thickness measurements are made at the Sandia Ion Beam Laboratory using Rutherford backscattering spectrometry and nuclear reaction analysis, as described in this report. This report describes the film deposition and pre-exposure analysis. Results from analysis after plasma exposure will be given in a subsequent report.

  1. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  2. Development of Operational Parameters for Advanced Voloxidation Process at KAERI

    International Nuclear Information System (INIS)

    KAERI has been developing a voloxidation process as a head-end process of pyroprocessing technology with INL (Idaho National Laboratory). The work scope of KAERI is to develop the operation parameters for advanced voloxidation process at KAERI using surrogate materials and SIMFUEL. In order to evaluate operation conditions of an advanced voloxidation process, oxidation and vaporization behavior of metals and Cs compounds was investigated in terms of thermal treatment atmosphere and temperature by using thermodynamic data. And also, the oxidation and vaporization behavior of semi-volatile fission products with process pressure and temperature was investigated using surrogate materials. Particle size control for U3O8 powder was investigated using SIMFUEL and a rotary voloxidizer. According to analysis of KAERI works, the operation conditions for advanced voloxiation process may be consisted of the following four steps: 1) oxidation of UO2 pellet into U3O8 powder at 500 .deg. C in oxidative atmosphere, 2) additional oxidation of noble metal alloy and vaporization of high vapor pressure of fission products at 700 .deg. C in oxidative atmosphere, 3) granulation of U3O8 powder and vaporization of Cs compounds at 1200 .deg. C in an atmosphere of argon, and 4) reduction of UO2+x granules into UO2 granules at 1000 .deg. C in an atmosphere of 4%H2-Ar. This report will be used as a useful means for determining the operation parameters for advanced voloxidation process

  3. Construction and initial operation of the Advanced Toroidal Facility

    International Nuclear Information System (INIS)

    The Advanced Toroidal Facility (ATF) torsatron was designed on a physics basis for access to the second stability regime and on an engineering basis for independent fabrication of high-accuracy components. The actual construction, assembly, and initial operation of ATF are compared with the characteristics expected during the design of ATF. 31 refs., 19 figs., 2 tabs

  4. Comparison of Windows and Linux Operating Systems in Advanced Features

    Directory of Open Access Journals (Sweden)

    P. Abhilash

    2015-02-01

    Full Text Available Comparison between the Microsoft Windows and Linux computer operating systems is a long-running discussion topic within the personal computer industry .This technical paper is mainly going to focus on the differences between windows and linux in all fields. Both Windows and Linux Operating systems have their own advantages and differ in functionalities and user friendliness. Linux and Microsoft Windows differ in philosophy, cost, versatility and stability, with each seeking to improve in their perceived weaker areas. This paper is mainly going to focus on the advanced features that are uniquely present in one operating system and not in other one.

  5. Ion cyclotron system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H

    1998-05-01

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs.

  6. Ion cyclotron system design for KSTAR tokamak

    International Nuclear Information System (INIS)

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs

  7. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  8. Operational Strategy of CBPs for load balancing of Operators in Advanced Main Control Room

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seunghwan; Kim, Yochan; Jung, Wondea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    With the using of a computer-based control room in an APR1400 (Advanced Pressurized Reactor-1400), the operators' behaviors in the main control room had changed. However, though the working environment of operators has been changed a great deal, digitalized interfaces can also change the cognitive tasks or activities of operators. First, a shift supervisor (SS) can confirm/check the conduction of the procedures and the execution of actions of board operators (BOs) while confirming directly the operation variables without relying on the BOs. Second, all operators added to their work the use of a new CBP and Soft Controls, increasing their procedural workload. New operational control strategies of CBPs are necessary for load balancing of operator's task load in APR1400. In this paper, we compared the workloads of operators in an APR1400 who work with two different usages of the CBP. They are SS oriented usage and SS-BO collaborative usage. In this research, we evaluated the workloads of operators in an advanced main control room by the COCOA method. Two types of CBP usages were defined and the effects of these usages on the workloads were investigated. The obtained results showed that the workloads between operators in a control room can be balanced according to the CBP usages by assigning control authority to the operators.

  9. Advanced Transport Operating System (ATOPS) control display unit software description

    Science.gov (United States)

    Slominski, Christopher J.; Parks, Mark A.; Debure, Kelly R.; Heaphy, William J.

    1992-01-01

    The software created for the Control Display Units (CDUs), used for the Advanced Transport Operating Systems (ATOPS) project, on the Transport Systems Research Vehicle (TSRV) is described. Module descriptions are presented in a standardized format which contains module purpose, calling sequence, a detailed description, and global references. The global reference section includes subroutines, functions, and common variables referenced by a particular module. The CDUs, one for the pilot and one for the copilot, are used for flight management purposes. Operations performed with the CDU affects the aircraft's guidance, navigation, and display software.

  10. ADVANCED COMPRESSOR ENGINE CONTROLS TO ENHANCE OPERATION, RELIABILITY AND INTEGRITY

    Energy Technology Data Exchange (ETDEWEB)

    Gary D. Bourn; Jess W. Gingrich; Jack A. Smith

    2004-03-01

    This document is the final report for the ''Advanced Compressor Engine Controls to Enhance Operation, Reliability, and Integrity'' project. SwRI conducted this project for DOE in conjunction with Cooper Compression, under DOE contract number DE-FC26-03NT41859. This report addresses an investigation of engine controls for integral compressor engines and the development of control strategies that implement closed-loop NOX emissions feedback.

  11. Advanced Transport Operating System (ATOPS) utility library software description

    Science.gov (United States)

    Clinedinst, Winston C.; Slominski, Christopher J.; Dickson, Richard W.; Wolverton, David A.

    1993-01-01

    The individual software processes used in the flight computers on-board the Advanced Transport Operating System (ATOPS) aircraft have many common functional elements. A library of commonly used software modules was created for general uses among the processes. The library includes modules for mathematical computations, data formatting, system database interfacing, and condition handling. The modules available in the library and their associated calling requirements are described.

  12. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  13. The design of the Tokamak Physics Experiment (TPX)

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, J.A.; Goldston, R.J.; Sinnis, J.C.; Bernabei, S.; Bialek, J.M.; Bronner, G.; Chen, S.J.; Chrzanowski, J.; Citrolo, J.; Dahlgren, F.

    1993-09-01

    The Tokamak Physics Experiment (TPX) is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a low-activation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation of US industry. Operation will begin with first plasma in the year 2000.

  14. Advanced Communications Technology Satellite Now Operating in an Inclined Orbit

    Science.gov (United States)

    Bauer, Robert A.

    1999-01-01

    The Advanced Communications Technology Satellite (ACTS) system has been modified to support operation in an inclined orbit that is virtually transparent to users, and plans are to continue this final phase of its operation through September 2000. The next 2 years of ACTS will provide a new opportunity for using the technologies that this system brought online over 5 years ago and that are still being used to resolve the technical issues that face NASA and the satellite industry in the area of seamless networking and interoperability with terrestrial systems. New goals for ACTS have been defined that align the program with recent changes in NASA and industry. ACTS will be used as a testbed to: Show how NASA and other Government agencies can use commercial systems for 1. future support of their operations Test, characterize, and resolve technical issues in using advanced communications 2. protocols such as asynchronous transfer mode (ATM) and transmission control protocol/Internet protocol (TCP/IP) over long latency links as found when interoperating satellites with terrestrial systems Evaluate narrow-spot-beam Ka-band satellite operation in an inclined orbit 3. Verify Ka-band satellite technologies since no other Ka-band system is yet 4. available in the United States

  15. Recent Progress of HT-7U Superconducting Tokamak

    Science.gov (United States)

    Weng, Pei-de

    2002-12-01

    HT-7U is a superconducting tokamak, which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak. The engineering design of the device has been optimized. The R&D program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.

  16. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (ne and Te) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  17. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System.

  18. Advancing reservoir operation description in physically based hydrological models

    Science.gov (United States)

    Anghileri, Daniela; Giudici, Federico; Castelletti, Andrea; Burlando, Paolo

    2016-04-01

    Last decades have seen significant advances in our capacity of characterizing and reproducing hydrological processes within physically based models. Yet, when the human component is considered (e.g. reservoirs, water distribution systems), the associated decisions are generally modeled with very simplistic rules, which might underperform in reproducing the actual operators' behaviour on a daily or sub-daily basis. For example, reservoir operations are usually described by a target-level rule curve, which represents the level that the reservoir should track during normal operating conditions. The associated release decision is determined by the current state of the reservoir relative to the rule curve. This modeling approach can reasonably reproduce the seasonal water volume shift due to reservoir operation. Still, it cannot capture more complex decision making processes in response, e.g., to the fluctuations of energy prices and demands, the temporal unavailability of power plants or varying amount of snow accumulated in the basin. In this work, we link a physically explicit hydrological model with detailed hydropower behavioural models describing the decision making process by the dam operator. In particular, we consider two categories of behavioural models: explicit or rule-based behavioural models, where reservoir operating rules are empirically inferred from observational data, and implicit or optimization based behavioural models, where, following a normative economic approach, the decision maker is represented as a rational agent maximising a utility function. We compare these two alternate modelling approaches on the real-world water system of Lake Como catchment in the Italian Alps. The water system is characterized by the presence of 18 artificial hydropower reservoirs generating almost 13% of the Italian hydropower production. Results show to which extent the hydrological regime in the catchment is affected by different behavioural models and reservoir

  19. Co-Operative Advances in Behavioral Health and Performance Research and Operations

    Science.gov (United States)

    VanderArk, Stephen T.; Leveton, Lauren B.

    2011-01-01

    In organizations that engage in both operations and applied research, with operational needs guiding research questions and research informing improved operations, the ideal goal is a synergy of ideas and information. In reality, this ideal synergy is often lacking. Real-time operational needs driving day-to-day decisions, lack of communication, lag time in getting research advances plugged into operations can cause both areas to suffer from this gap between operations and research. At Johnson Space Center, the Behavior Health and Performance group (BHP) strives to bridge this gap by following a Human Research Program framework: Expectations of future operational needs identify the knowledge gaps; the gaps in turn guide research leading to a product that is transitioned into operations. Thus, the direction those of us in research take is in direct response to current and future needs of operations. Likewise, those of us in operations actively seek knowledge that is supported by evidence-based research. We make an ongoing effort to communicate across the research and operations gap by working closely with each other and making a conscious effort to keep each other informed. The objective of the proposed panel discussion is to demonstrate through the following presentations the results of a successful collaboration between research and operations and to provide ASMA members with more practical knowledge and strategies for building these bridges to serve our field of practice well. The panel will consist of six presenters from BHP operations, internal BHP research, and external research instigated by BHP who together represent the entire BHP Research Transition to Operations Framework

  20. Assessment of the roles of the Advanced Neutron Source Operators

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is unique in the extent to which human factors engineering (HFE) principles are being applied at the conceptual design stage. initial HFE accomplishments include the development of an ANS HFE program plan, operating philosophy, and functional analysis. In FY 1994, HFE activities focused on the role of the ANS control room reactor operator (RO). An operator-centered control room model was used in conjunction with information gathered from existing ANS system design descriptions and other literature to define a list of RO responsibilities. From this list, a survey instrument was developed and administered to ANS design engineers, operations management personnel at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR), and HFIR ROs to detail the nature of the RO position. Initial results indicated that the RO will function as a high-level system supervisor with considerable monitoring, verification, and communication responsibilities. The relatively high level of control automation has resulted in a reshaping of the RO's traditional safety and investment protection roles

  1. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  2. Preliminary R and D on flat-type W/Cu plasma-facing materials and components for Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    To upgrade the Experimental Advanced Superconducting Tokamak dome and first-wall, flat-type W/Cu plasma-facing components will be installed in the coming years in order to exhaust the increasing heat flux. Mock-ups with an interlayer of oxygen-free Cu (OFC) made by vacuum hot pressing have been developed and the bonding strength was found to be over 100 MPa. The behavior of the mock-ups under steady-state high heat flux loads has been studied. No crack or exfoliation occurred on the W surface and W/OFC/CuCrZr interfaces after screening tests with heat fluxes of 2.24–7.73 MW m−2. The mock-up survived up to 1000 cycles heat load of 3.24 MW m−2 with cooling water of 4 m s−1, 20 °C. However, cracks appeared in W around the gaps at about the 300th cycle under a heat load of 5.37 MW m−2. We have also studied the chemical vapor deposition W coated CuCrZr with an OFC interlayer. It has been found that: (i) the OFC interlayer plays a significant role in achieving coatings without any crack, (ii) the deposition rate was about 0.3–0.5 mm h−1 at 490–580 °C and (iii) a bonding strength of 53.7 MPa was achieved with laser surfi-sculpt. (paper)

  3. Integrated Refrigeration and Storage for Advanced Liquid Hydrogen Operations

    Science.gov (United States)

    Swanger, A. M.; Notardonato, W. U.; Johnson, W. L.; Tomsik, T. M.

    2016-01-01

    NASA has used liquefied hydrogen (LH2) on a large scale since the beginning of the space program as fuel for the Centaur and Apollo upper stages, and more recently to feed the three space shuttle main engines. The LH2 systems currently in place at the Kennedy Space Center (KSC) launch pads are aging and inefficient compared to the state-of-the-art. Therefore, the need exists to explore advanced technologies and operations that can drive commodity costs down, and provide increased capabilities. The Ground Operations Demonstration Unit for Liquid Hydrogen (GODU-LH2) was developed at KSC to pursue these goals by demonstrating active thermal control of the propellant state by direct removal of heat using a cryocooler. The project has multiple objectives including zero loss storage and transfer, liquefaction of gaseous hydrogen, and densification of liquid hydrogen. The key technology challenge was efficiently integrating the cryogenic refrigerator into the LH2 storage tank. A Linde LR1620 Brayton cycle refrigerator is used to produce up to 900W cooling at 20K, circulating approximately 22 g/s gaseous helium through the hydrogen via approximately 300 m of heat exchanger tubing. The GODU-LH2 system is fully operational, and is currently under test. This paper will discuss the design features of the refrigerator and storage system, as well as the current test results.

  4. Advanced EMS and its trial operation in Shanghai power system

    Institute of Scientific and Technical Information of China (English)

    LU Qiang; HE GuangYu; MEI ShengWei; SUN YingYun; RUAN QianTu; WANG Wei; ZHANG WangJun; YU XuFeng

    2008-01-01

    To meet the demand of high stability, high quality, and low losses of power systems, the advanced energy management system (AEMS) is established and revealed in this bulletin, which has been put into trial operation in Shanghai power system for almost half a year. The AEMS is novel from all aspects covering idea, theory, method, software, and engineering. The essence of AEMS is exercising the hybrid automatic control theory and technology to realize multi-objective optimal closedloop control of power systems. Based on an "event-driven" strategy, the AEMS transforms multi-objective optimal control problems into event identification and elimination by defining the unsatisfactory states of a power system as events. This bulletin concisely presents the theory and main advantages of AEMS, as well as its implementation in Shanghai power system.

  5. Advanced EMS and its trial operation in Shanghai power system

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    To meet the demand of high stability,high quality,and low losses of power systems,the advanced energy management system (AEMS) is established and revealed in this bulletin,which has been put into trial operation in Shanghai power system for almost half a year. The AEMS is novel from all aspects covering idea,theory,method,software,and engineering. The essence of AEMS is exercising the hybrid automatic control theory and technology to realize multi-objective optimal closed-loop control of power systems. Based on an "event-driven" strategy,the AEMS transforms multi-objective optimal control problems into event identification and elimination by defining the unsatisfactory states of a power system as events. This bulletin concisely presents the theory and main advantages of AEMS,as well as its implementation in Shanghai power system.

  6. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  7. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  8. Tokamak Systems Code

    International Nuclear Information System (INIS)

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  9. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

    2002-07-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  10. TOKAMAK-15 modernization and an analysis of cryogenic system operation for the period from 1988 to 1994

    International Nuclear Information System (INIS)

    The T-15 cryogenics system has been designed for cooling down, cryostatting, warming up of superconducting, cryoresistive and cryogenics T-15 objects. Maintenance of the cryogenics system has been on going since 1988. For the mentioned period, in the cryogenics T-15 system. The capacity of screw compressor was increased from 0.181 kg/s to 0.236 kg/s (third stage compressors with increased capacity were developed and manufactured), their reliability was also enhanced. The capacity of liquefiers was increased from 0.0833 - 0.0972 L/s (300-350 L/h) to 0.222 L/s (800 L/h) due to replacement of turboexpanders by more effective ones and due to introduction of an end-stage turboexpander into maintenance. The heat influxes to the cryogenics pipelines were reduced by 50%. For the same period some technological regimes of cryogenics system have been developed to produce the maximal output of cold. The cooling down from 110 K to 15 K is done, when one or two liquefiers are in operation under refrigerating conditions with the reverse flow splitting. The further cooling is performed under joint operation of two liquefiers; one of them operates in the liquefying mode, another, in the refrigerating one with excess reverse flow. A change in the operating conditions was necessary because of the impossibility of regulating the distribution of the reverse helium flow between two liquefiers at the temperature below 15K. The main regime at the level of 4.5 K is a two-loop operating diagram, when one liquefier and a passive refrigerator with excessive reverse flow are in operation, the refrigerating capacity is about 3 kW

  11. Research using small tokamaks

    International Nuclear Information System (INIS)

    The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway

  12. Dust Measurements in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  13. Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

    International Nuclear Information System (INIS)

    SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

  14. First experiments on the TO-2 tokamak with a divertor

    International Nuclear Information System (INIS)

    Long stable discharges have been obtained in a recetrack tokamak with toroidal divertors in low plasma density regime. Divertors sharply limit plasma filament cross section, plasma density decreasing by an order at 1 cm length near the separatrix. 8 mm thick well formed flux of plasma appears at the divertor plate. Divertor power efficiency at different modes of operation is 50- 70 %. As compared to the TO-1 nondivertor tokamak some plasma filament hot zone expansion is recorded in the TO-2 tokamak

  15. Analysis of neutral hydrogenic emission spectra in a tokamak

    Science.gov (United States)

    Ko, J.; Chung, J.; Jaspers, R. J. E.

    2015-10-01

    Balmer-α radiation by the excitation of thermal and fast neutral hydrogenic particles has been investigated in a magnetically confined fusion device, or tokamak, from the Korea Superconducting Tokamak Advanced Research (KSTAR). From the diagnostic point of view, the emission from thermal neutrals is associated with passive spectroscopy and that from energetic neutrals that are usually injected from the outside of the tokamak to the active spectroscopy. The passive spectroscopic measurement for the thermal Balmer-α emission from deuterium and hydrogen estimates the relative concentration of hydrogen in a deuterium-fueled plasma and therefore, makes a useful tool to monitor the vacuum wall condition. The ratio of hydrogen to deuterium obtained from this measurement qualitatively correlates with the energy confinement of the plasma. The Doppler-shifted Balmer-α components from the fast neutrals features the spectrum of the motional Stark effect (MSE) which is an essential principle for the measurement of the magnetic pitch angle profile. Characterization of this active MSE spectra, especially with multiple neutral beam lines crossing along the observation line of sight, has been done for the guideline of the multi-ion-source heating beam operation and for the optimization of the narrow bandpass filters that are required for the polarimeter-based MSE diagnostic system under construction at KSTAR.

  16. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  17. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    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

  18. Bootstrap Current in Spherical Tokamaks

    Institute of Scientific and Technical Information of China (English)

    王中天; 王龙

    2003-01-01

    Variational principle for the neoclassical theory has been developed by including amomentum restoring term in the electron-electron collisional operator, which gives an additionalfree parameter maximizing the heat production rate. All transport coefficients are obtained in-cluding the bootstrap current. The essential feature of the study is that the aspect ratio affects thefunction of the electron-electron collision operator through a geometrical factor. When the aspectratio approaches to unity, the fraction of circulating particles goes to zero and the contribution toparticle flux from the electron-electron collision vanishes. The resulting diffusion coefficient is inrough agreement with Hazeltine. When the aspect ratio approaches to infinity, the results are inagreement with Rosenbluth. The formalism gives the two extreme cases a connection. The theoryis particularly important for the calculation of bootstrap current in spherical tokamaks and thepresent tokamaks, in which the square root of the inverse aspect ratio, in general, is not small.

  19. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  20. Designing a tokamak fusion reactor—How does plasma physics fit in?

    Science.gov (United States)

    Freidberg, J. P.; Mangiarotti, F. J.; Minervini, J.

    2015-07-01

    This paper attempts to bridge the gap between tokamak reactor design and plasma physics. The analysis demonstrates that the overall design of a tokamak fusion reactor is determined almost entirely by the constraints imposed by nuclear physics and fusion engineering. Virtually, no plasma physics is required to determine the main design parameters of a reactor: a , R 0 , B 0 , T i , T e , p , n , τ E , I . The one exception is the value of the toroidal current I , which depends upon a combination of engineering and plasma physics. This exception, however, ultimately has a major impact on the feasibility of an attractive tokamak reactor. The analysis shows that the engineering/nuclear physics design makes demands on the plasma physics that must be satisfied in order to generate power. These demands are substituted into the well-known operational constraints arising in tokamak physics: the Troyon limit, Greenwald limit, kink stability limit, and bootstrap fraction limit. Unfortunately, a tokamak reactor designed on the basis of standard engineering and nuclear physics constraints does not scale to a reactor. Too much current is required to achieve the necessary confinement time for ignition. The combination of achievable bootstrap current plus current drive is not sufficient to generate the current demanded by the engineering design. Several possible solutions are discussed in detail involving advances in plasma physics or engineering. The main contribution of the present work is to demonstrate that the basic reactor design and its plasma physics consequences can be determined simply and analytically. The analysis thus provides a crisp, compact, logical framework that will hopefully lead to improved physical intuition for connecting plasma physic to tokamak reactor design.

  1. Advanced Materials for RSOFC Dual Operation with Low Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Eric, Tang; Tony, Wood; Sofiane, Benhaddad; Casey, Brown; Hongpeng, He; Jeff, Nelson; Oliver, Grande; Ben, Nuttall; Mark, Richards; Randy, Petri

    2012-12-27

    Reversible solid oxide fuel cells (RSOFCs) are energy conversion devices. They are capable of operating in both power generation mode (SOFC) and electrolysis modes (SOEC). RSOFC can integrate renewable production of electricity and hydrogen when power generation and steam electrolysis are coupled in a system, which can turn intermittent solar and wind energy into "firm power." In this DOE EERE project, VPS continuously advanced RSOFC cell stack technology in the areas of endurance and performance. Over 20 types of RSOFC cells were developed in the project. Many of those exceeded performance (area specific resistance less than 300 mohmcm2) and endurance (degradation rate less than 4% per 1000 hours) targets in both fuel cell and electrolysis modes at 750C. One of those cells, RSOFC-7, further demonstrated the following: Steady-state electrolysis with a degradation rate of 1.5% per 1000 hours. Ultra high current electrolysis over 3 A/cm2 at 75% water electrolysis efficiency voltage of 1.67 V. Daily SOFC/SOEC cyclic test of over 600 days with a degradation rate of 1.5% per 1000 hours. Over 6000 SOFC/SOEC cycles in an accelerated 20-minute cycling with degradation less than 3% per 1000 cycles. In RSOFC stack development, a number of kW-class RSOFC stacks were developed and demonstrated the following: Steady-state electrolysis operation of over 5000 hours. Daily SOFC/SOEC cyclic test of 100 cycles. Scale up capability of using large area cells with 550 cm2 active area showing the potential for large-scale RSOFC stack development in the future. Although this project is an open-ended development project, this effort, leveraging Versa Power Systems' years of development experience, has the potential to bring renewable energy RSOFC storage systems significantly closer to commercial viability through improvements in RSOFC durability, performance, and cost. When unitized and deployed in renewable solar and wind installations, an RSOFC system can enable higher

  2. Experimental data base of Tokamak KTM physical diagnostics

    International Nuclear Information System (INIS)

    The process of software creation of experimental data storage of Tokamak KTM physical diagnostics based on analysis of storage methods of operating Tokamaks data is considered. Task of specific kinds of information storage is solved; experimental data base that is thr part of system providing information analysis performance in the post-start period is developed.(author)

  3. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  4. Reference Operational Concepts for Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques Victor [Idaho National Lab. (INL), Idaho Falls, ID (United States); Farris, Ronald Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report represents the culmination of a four-year research project that was part of the Instrumentation and Control and Human Machine Interface subprogram of the DOE Advanced Reactor Technologies program.

  5. Operator’s cognitive, communicative and operative activities based workload measurement of advanced main control room

    International Nuclear Information System (INIS)

    Highlights: • An advanced MMIS in the advanced MCR requires new roles and tasks of operators. • A new workload evaluation framework is needed for a new MMIS environment. • This work suggests a new workload measurement approach (COCOA) for an advanced MCR. • COCOA enables 3-dimensional measurement of cognition, communication and operation. • COCOA workload evaluation of the reference plant through simulation was performed. - Abstract: An advanced man–machine interface system (MMIS) with a computer-based procedure system and high-tech control/alarm system is installed in the advanced main control room (MCR) of a nuclear power plant. Accordingly, though the task of the operators has been changed a great deal, owing to a lack of appropriate guidelines on the role allocation or communication method of the operators, operators should follow the operating strategies of conventional MCR and the problem of an unbalanced workload for each operator can be raised. Thus, it is necessary to enhance the operation capability and improve the plant safety by developing guidelines on the role definition and communication of operators in an advanced MCR. To resolve this problem, however, a method for measuring the workload according to the work execution of the operators is needed, but an applicable method is not available. In this research, we propose a COgnitive, Communicative and Operational Activities measurement approach (COCOA) to measure and evaluate the workload of operators in an advanced MCR. This paper presents the taxonomy for additional operation activities of the operators to use the computerized procedures and soft control added to an advanced MCR, which enables an integrated measurement of the operator workload in various dimensions of cognition, communication, and operation. To check the applicability of COCOA, we evaluated the operator workload of an advanced MCR of a reference power plant through simulation training experiments. As a result, the amount

  6. The Experiments of the small Spherical Tokamak Gutta

    International Nuclear Information System (INIS)

    GUTTA is a small spherical tokamak (R = 16cm, a = 8cm, Ip = 150kA) operating at the St. Petersburg State University since 2004 in the scope of the IAEA CRP ''Joint Research using Small Tokamaks''. Main scientific activities on GUTTA include development of new and improvement of existing mathematical models of plasma control, relevant for application on large tokamaks and ITER and verification of them on GUTTA; studies on the ECRH/EBW assisted breakdown and non-solenoid plasma formation in low aspect ratio tokamak; development of diagnostics; training and education of students.In this paper design properties of Gutta will be presented. Regimes of operation of the tokamak and plasma shape parameters are described and first results of the plasma formation and start-up studied will be discussed

  7. Strength advance design of boiler components for DSS operation

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Tsuyoshi; Tsuta, Toshio; Yamaji, Seiichi; Miyoshi, Teiichi (Kawasaki Heavy Industries Ltd., Kobe, (Japan))

    1989-08-20

    The thermal power plants in the future are expected to operate under such medium loads as LNG, petroleum and coal burning. As a result, frequent start-ups at night (DSS operation) and week-end stops (WSS operation), etc. are conducted: sliding pressure operation system is adopted to reduce the thermal stress at the turbine and to improve the plant efficiency at partial load operation; this causes a rapid temperture change which gives rise to big thermal stress at various points of the boiler causing wide fatigue damage at the center of the stress. Simultneously, the creep damage at high temperature progresses by time. In order to attain sufficient life for the planned operation mode at the design of various boiler points, improvements have been conducted at internal bore edge of a thick cylinder, shape of the fin end piece, furnace/horizontal flue corner part and the attached metal pieces for enhancing the reliability. 1 ref., 9 figs., 1 tab.

  8. Advances in Operational Flood Risk Management in the Netherlands

    NARCIS (Netherlands)

    Wojciechowska, K.A.

    2015-01-01

    Operational flood risk management refers to activities that aim to reduce the probability and/or negative consequences of flooding just prior to the expected flood event. An inherent feature of operational flood risk management is that outcomes of decisions taken are uncertain. The main goal of this

  9. Optimization of advanced plants operation: The Escrime project

    International Nuclear Information System (INIS)

    The Escrime program aims at defining the optimal share of tasks between humans and computers under normal or accidental plant operation. Basic principles we keep in mind are the following: human operators are likely to be necessary in the operation of future plants because we cannot demonstrate that plant design is error free, so unexpected situation can still happen; automation must not release the operators from their decisional role but only help them avoiding situations of cognitive overload which can lead to increase the risk of errors; the optimum share of tasks between human and automatic systems must be based on a critical analysis of the tasks and of the way they are handled. The last point appeared to be of major importance. The corresponding analysis of the French PWR's operating procedures enabled us to define a unified scheme for plant operation under the form of a hierarchy of goals and means. Beyond this analysis, development of a specific testing facility is under way to check the relevance of the proposed plant operation organization and to test the human-machine cooperation in different situations for various levels of automation. 7 refs, 4 figs

  10. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  11. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    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

  12. ESCRIME: testing bench for advanced operator workstations in future plants

    International Nuclear Information System (INIS)

    The problem of optimal task allocation between man and computer for the operation of nuclear power plants is of major concern for the design of future plants. As the increased level of automation induces the modification of the tasks actually devoted to the operator in the control room, it is very important to anticipate these consequences at the plant design stage. The improvement of man machine cooperation is expected to play a major role in minimizing the impact of human errors on plant safety. The CEA has launched a research program concerning the evolution of the plant operation in order to optimize the efficiency of the human/computer systems for a better safety. The objective of this program is to evaluate different modalities of man-machine share of tasks, in a representative context. It relies strongly upon the development of a specific testing facility, the ESCRIME work bench, which is presented in this paper. It consists of an EDF 1300MWe PWR plant simulator connected to an operator workstation. The plant simulator model presents at a significant level of details the instrumentation and control of the plant and the main connected circuits. The operator interface is based on the generalization of the use of interactive graphic displays, and is intended to be consistent to the tasks to be performed by the operator. The functional architecture of the workstation is modular, so that different cooperation mechanisms can be implemented within the same framework. It is based on a thorough analysis and structuration of plant control tasks, in normal as well as in accident situations. The software architecture design follows the distributed artificial intelligence approach. Cognitive agents cooperate in order to operate the process. The paper presents the basic principles and the functional architecture of the test bed and describes the steps and the present status of the program. (author)

  13. Advanced construction management for lunar base construction - Surface operations planner

    Science.gov (United States)

    Kehoe, Robert P.

    1992-01-01

    The study proposes a conceptual solution and lays the framework for developing a new, sophisticated and intelligent tool for a lunar base construction crew to use. This concept integrates expert systems for critical decision making, virtual reality for training, logistics and laydown optimization, automated productivity measurements, and an advanced scheduling tool to form a unique new planning tool. The concept features extensive use of computers and expert systems software to support the actual work, while allowing the crew to control the project from the lunar surface. Consideration is given to a logistics data base, laydown area management, flexible critical progress scheduler, video simulation of assembly tasks, and assembly information and tracking documentation.

  14. Characterization of the Tokamak Novillo in cleaning regime; Caracterizacion del Tokamak Novillo en regimen de limpieza

    Energy Technology Data Exchange (ETDEWEB)

    Lopez C, R.; Melendez L, L.; Valencia A, R.; Chavez A, E.; Colunga S, S.; Gaytan G, E

    1992-02-15

    In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip{sub t} like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I{sub (p)}t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)

  15. Real-time software for the COMPASS tokamak plasma control

    Energy Technology Data Exchange (ETDEWEB)

    Valcarcel, D.F., E-mail: danielv@ipfn.ist.utl.p [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic)

    2010-07-15

    The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 {mu}s. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.

  16. Multichannel submillimeter interferometer for tokamak density measurements

    International Nuclear Information System (INIS)

    A two-channel, submillimeter (SMM) laser, electron-density interferometer has been operated successfully on the ISX tokamak. The interferometer is the first phase of a diagnostic system to measure the tokamak plasma current density using the Faraday rotation of the polarization vector of SMM laser beams. Deuterated formic acid lasers (lambda = 0.381 mm) have produced cw power of 10 mW. The interferometer has performed successfully for line-averaged electron densities as high as 8 x 1013 cm-3

  17. Centrifugal microfluidic platforms: advanced unit operations and applications.

    Science.gov (United States)

    Strohmeier, O; Keller, M; Schwemmer, F; Zehnle, S; Mark, D; von Stetten, F; Zengerle, R; Paust, N

    2015-10-01

    Centrifugal microfluidics has evolved into a mature technology. Several major diagnostic companies either have products on the market or are currently evaluating centrifugal microfluidics for product development. The fields of application are widespread and include clinical chemistry, immunodiagnostics and protein analysis, cell handling, molecular diagnostics, as well as food, water, and soil analysis. Nevertheless, new fluidic functions and applications that expand the possibilities of centrifugal microfluidics are being introduced at a high pace. In this review, we first present an up-to-date comprehensive overview of centrifugal microfluidic unit operations. Then, we introduce the term "process chain" to review how these unit operations can be combined for the automation of laboratory workflows. Such aggregation of basic functionalities enables efficient fluidic design at a higher level of integration. Furthermore, we analyze how novel, ground-breaking unit operations may foster the integration of more complex applications. Among these are the storage of pneumatic energy to realize complex switching sequences or to pump liquids radially inward, as well as the complete pre-storage and release of reagents. In this context, centrifugal microfluidics provides major advantages over other microfluidic actuation principles: the pulse-free inertial liquid propulsion provided by centrifugal microfluidics allows for closed fluidic systems that are free of any interfaces to external pumps. Processed volumes are easily scalable from nanoliters to milliliters. Volume forces can be adjusted by rotation and thus, even for very small volumes, surface forces may easily be overcome in the centrifugal gravity field which enables the efficient separation of nanoliter volumes from channels, chambers or sensor matrixes as well as the removal of any disturbing bubbles. In summary, centrifugal microfluidics takes advantage of a comprehensive set of fluidic unit operations such as

  18. Research using small tokamaks

    International Nuclear Information System (INIS)

    These proceedings of the IAEA-sponsored meeting held in Nice, France 10-11 October, 1988, contain the manuscripts of the 21 reports dealing with research using small tokamaks. The purpose of this meeting was to highlight some of the achievements of small tokamaks and alternative magnetic confinement concepts and assess the suitability of starting new programs, particularly in developing countries. Papers presented were either review papers, or were detailed descriptions of particular experiments or concepts. Refs, figs and tabs

  19. The conceptual design of a robust, compact, modular tokamak reactor based on high-field superconductors

    Science.gov (United States)

    Whyte, D. G.; Bonoli, P.; Barnard, H.; Haakonsen, C.; Hartwig, Z.; Kasten, C.; Palmer, T.; Sung, C.; Sutherland, D.; Bromberg, L.; Mangiarotti, F.; Goh, J.; Sorbom, B.; Sierchio, J.; Ball, J.; Greenwald, M.; Olynyk, G.; Minervini, J.

    2012-10-01

    Two of the greatest challenges to tokamak reactors are 1) large single-unit cost of each reactor's construction and 2) their susceptibility to disruptions from operation at or above operational limits. We present an attractive tokamak reactor design that substantially lessens these issues by exploiting recent advancements in superconductor (SC) tapes allowing peak field on SC coil > 20 Tesla. A R˜3.3 m, B˜9.2 T, ˜ 500 MW fusion power tokamak provides high fusion gain while avoiding all disruptive operating boundaries (no-wall beta, kink, and density limits). Robust steady-state core scenarios are obtained by exploiting the synergy of high field, compact size and ideal efficiency current drive using high-field side launch of Lower Hybrid waves. The design features a completely modular replacement of internal solid components enabled by the demountability of the coils/tapes and the use of an immersion liquid blanket. This modularity opens up the possibility of using the device as a nuclear component test facility.

  20. Mass spectrometry instrumentation in TN (Novillo Tokamak)

    International Nuclear Information System (INIS)

    The mass spectrophotometry in the residual gases analysis in high vacuum systems, in particular in the Novillo Tokamak (TN), where pressures are required to be of the order 10-7 Torr, is carried out through an instrumental support with infrastructure configured in parallel to the experimental planning in this device. In the Novillo as well as other Tokamaks, it is necessary to condition the vacuum chamber for improving the main discharge parameters. At the present time, in this Tokamak the conditioning quality is presented determined by means of a mass spectrophotometer. A general instrumental description is presented associated with the Novillo conditioning, as well as the spectras obtained before and after operation. (Author)

  1. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  2. Recent advances in TEC-less uncooled FPA sensor operation

    Science.gov (United States)

    Howard, Philip E.; Clarke, John E.; Li, Chuan C.; Yang, John W.; Wong, W. Y.; Bogosyan, Arsen

    2003-09-01

    DRS has previously demonstrated and reported a concept for operating uncooled infrared focal plane arrays (UIRFPA) without the need for UIRFPA temperature regulation. DRS has patented this proprietary technology, which DRS calls TCOMP. TCOMP is a concept that combines an operating algorithm, a sensor architecture and a sensor calibration method, which allow pixel response and offset correction to be performed as a function of the UFPA sensor's operating temperature, thereby eliminating the need for the UIRFPA temperature regulation that would be required otherwise. As a result of the elimination of the temperature regulation requirement, the sensor turn-on time for high performance imaging can be significantly reduced, sensor power is significantly reduced, and the need for stray thermal radiation shields is effectively eliminated. The original TCOMP technique was demonstrated in 1998. Since then DRS has made significant improvements in both the TCOMP algorithm and the calibration process. This paper describes the patented TCOMP concept, presents the results of analysis of the improved TCOMP concept, and provides sensor level data of UIRFPA/sensor performance with the improved TCOMP algorithm.

  3. Advanced materials and structures for extreme operating conditions

    CERN Document Server

    Skrzypek, Jacek J; Rustichelli, Franco

    2008-01-01

    Increasing industrial demands for high temperature applications, high t- perature gradients, high heat cycle resistance, high wear resistance, impact resistance, etc. , require application of new materials. Conventional met- lic materials, such as steels, nickel- and aluminium-based alloys, etc. c- not resist such extreme operating conditions. They have to be replaced by new metal/matrix or ceramic/matrix composite materials, MMC or CMC, such as titanium/zirconia,titanium/alumina, nickel/zirconia,nickel/alumina, steel/chromium nitride MMCs, or titanium carbide/silicon carbide, alumina/ zirconi

  4. Development of advanced automatic operation system for nuclear ship. 1. Perfect automatic normal operation

    Energy Technology Data Exchange (ETDEWEB)

    Nakazawa, Toshio; Yabuuti, Noriaki; Takahashi, Hiroki; Shimazaki, Junya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-02-01

    Development of operation support system such as automatic operating system and anomaly diagnosis systems of nuclear reactor is very important in practical nuclear ship because of a limited number of operators and severe conditions in which receiving support from others in a case of accident is very difficult. The goal of development of the operation support systems is to realize the perfect automatic control system in a series of normal operation from the reactor start-up to the shutdown. The automatic control system for the normal operation has been developed based on operating experiences of the first Japanese nuclear ship `Mutsu`. Automation technique was verified by `Mutsu` plant data at manual operation. Fully automatic control of start-up and shutdown operations was achieved by setting the desired value of operation and the limiting value of parameter fluctuation, and by making the operation program of the principal equipment such as the main coolant pump and the heaters. This report presents the automatic operation system developed for the start-up and the shutdown of reactor and the verification of the system using the Nuclear Ship Engineering Simulator System. (author)

  5. Advances in Service and Operations for ATLAS Data Management

    CERN Document Server

    Stewart, GA; The ATLAS collaboration

    2011-01-01

    ATLAS has recorded almost 5PB of RAW data since the LHC started running at the end of 2009. Many more derived data products and complimentary simulation data have also been produced by the collaboration and, in total, 55PB is currently stored in the Worldwide LHC Computing Grid by ATLAS. All of this data is managed by the ATLAS Distributed Data Management system, called Don Quixote 2 (DQ2). DQ2 has evolved rapidly to help ATLAS Computing operations to manage these large quantities of data across the many grid sites at which ATLAS runs and to help ATLAS physicists get access to this data. In this paper we describe new and improved DQ2 services: - Popularity service, which measures usage of data across ATLAS. - Space monitoring and accounting at sites. - Automated blacklisting service. - Cleaning agents, which trigger deletion of unused data at sites. - Deletion agents, to reliably delete unwanted data from sites. We describe the experience of data management operation in ATLAS computing, showing how these serv...

  6. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  7. ADVANCED, LOW/ZERO EMISSION BOILER DESIGN AND OPERATION

    Energy Technology Data Exchange (ETDEWEB)

    Fabienne Chatel-Pelage

    2003-10-01

    This document reviews the work performed during the quarter July--September 2003. Significant progress has been made in Task 1 (Site Preparation), Task 2 (Test performance) and Task 3 (Techno-Economic Study) of the project: the site preparation has been completed, two weeks of tests have been performed and the power generating units to be compared from an economical standpoint have been selected and accurately described. In the experimental part of this effort (task1), the partners in this project demonstrated the feasibility of 100% air replacement with O{sub 2}-enriched flue gas on 1.5MW coal-fired boiler. The air infiltration have been reduced to approximately 5% of the stoichiometry, enabling to reach around 70% of CO{sub 2} in the flue gases. Higher air in-leakage reduction is expected using alternative boiler operating procedure in order to achieve higher CO{sub 2} concentration in flue gas for further sequestration or reuse. The NO{sub x} emissions have been shown considerably lower in O{sub 2}-fired conditions than in air-baseline, the reduction rate averaging 70%. An additional week of tests is scheduled mid October 2003 for combustion parameter optimization, and some more days of operation will be dedicated to mercury emission measurement and heat transfer characterization. Out of the $485k already allocated in this project, $300k has been spent and reported to date, mainly in site preparation ({approx}$215k) and test performance ({approx}$85k). In addition to DOE allocated funds, to date approximately $240k has been cost-shared by the participants, bringing the total project cost up to $540k as on September 30, 2003.

  8. Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

    International Nuclear Information System (INIS)

    Tore-Supra has a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T proportional to √(t)) during the discharge until radiation cooling is equal to the incoming heat flux (T > 1800 K). Such an environment is found in most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra infrared thermography system is composed of 7 infrared endoscopes, this system is described in details in the paper, the new JET infrared thermography system is presented and some insights of the ITER set of visible/infrared endoscope is given. (authors)

  9. Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Guilhem, D.; Adjeroud, B.; Balorin, C.; Buravand, Y.; Bertrand, B.; Bondil, J.L.; Desgranges, C.; Gauthier, E.; Lipa, M.; Messina, P.; Missirlian, M.; Mitteau, R.; Moulin, D.; Pocheau, C.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.; Vallet, S

    2004-07-01

    Tore-Supra has a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T proportional to {radical}(t)) during the discharge until radiation cooling is equal to the incoming heat flux (T > 1800 K). Such an environment is found in most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra infrared thermography system is composed of 7 infrared endoscopes, this system is described in details in the paper, the new JET infrared thermography system is presented and some insights of the ITER set of visible/infrared endoscope is given. (authors)

  10. Advances in operational safety and severe accident research

    Energy Technology Data Exchange (ETDEWEB)

    Simola, K. [VTT Automation (Finland)

    2002-02-01

    A project on reactor safety was carried out as a part of the NKS programme during 1999-2001. The objective of the project was to obtain a shared Nordic view of certain key safety issues related to the operating nuclear power plants in Finland and Sweden. The focus of the project was on selected central aspects of nuclear reactor safety that are of common interest for the Nordic nuclear authorities, utilities and research bodies. The project consisted of three sub-projects. One of them concentrated on the problems related to risk-informed deci- sion making, especially on the uncertainties and incompleteness of probabilistic safety assessments and their impact on the possibilities to use the PSA results in decision making. Another sub-project dealt with questions related to maintenance, such as human and organisational factors in maintenance and maintenance management. The focus of the third sub-project was on severe accidents. This sub-project concentrated on phenomenological studies of hydrogen combustion, formation of organic iodine, and core re-criticality due to molten core coolant interaction in the lower head of reactor vessel. Moreover, the current status of severe accident research and management was reviewed. (au)

  11. Advances in Service and Operations for ATLAS Data Management

    CERN Document Server

    Stewart, G A; The ATLAS collaboration; Lassnig, M; Molfetas, A; Baristis, M; Zhang, D; Calvet, I; Beermann, T; Barreiro Megino, F; Tykhonov, A; Campana, S; Serfon, C; Oleynik, O; Petrosyan, A

    2012-01-01

    ATLAS has recorded almost 5PB of RAW data since the LHC started running at the end of 2009. Many more derived data products and complimentary simulation data have also been produced by the collaboration and, in total, 70PB is currently stored in the Worldwide LHC Computing Grid by ATLAS. All of this data is managed by the ATLAS Distributed Data Management system, called Don Quixote 2 (DQ2). DQ2 has evolved rapidly to help ATLAS Computing operations manage these large quantities of data across the many grid sites at which ATLAS runs and to help ATLAS physicists get access to this data. In this paper we describe new and improved DQ2 services: egin{itemize} item hspace{2mm} Popularity service, which measures usage of data across ATLAS. item hspace{2mm} Space monitoring and accounting at sites. item hspace{2mm} Automated exclusion service. item hspace{2mm} Cleaning agents, which trigger deletion of unused data at sites. item hspace{2mm} Deletion agents, to reliably delete unwanted data from sites. end{itemize} We...

  12. Advances in operational safety and severe accident research

    International Nuclear Information System (INIS)

    A project on reactor safety was carried out as a part of the NKS programme during 1999-2001. The objective of the project was to obtain a shared Nordic view of certain key safety issues related to the operating nuclear power plants in Finland and Sweden. The focus of the project was on selected central aspects of nuclear reactor safety that are of common interest for the Nordic nuclear authorities, utilities and research bodies. The project consisted of three sub-projects. One of them concentrated on the problems related to risk-informed deci- sion making, especially on the uncertainties and incompleteness of probabilistic safety assessments and their impact on the possibilities to use the PSA results in decision making. Another sub-project dealt with questions related to maintenance, such as human and organisational factors in maintenance and maintenance management. The focus of the third sub-project was on severe accidents. This sub-project concentrated on phenomenological studies of hydrogen combustion, formation of organic iodine, and core re-criticality due to molten core coolant interaction in the lower head of reactor vessel. Moreover, the current status of severe accident research and management was reviewed. (au)

  13. BEACON - An advanced continuous core monitoring and operational support system for pressurized water reactors

    International Nuclear Information System (INIS)

    An advanced continuous core monitoring and operational support system, BEACON, has been developed which combines a super fast nodal model, workstation based hardware, and existing instrumentation which can be used to improve plant availability and operating margin. (author). 6 refs, 8 figs

  14. The vacuum vessel thermal shield of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, B.J. E-mail: bjyoon@kaeri.re.kr; In, S.R.; Cho, S.Y

    2003-09-01

    The Korea superconducting tokamak advanced research (KSTAR) tokamak has an all-superconductor magnet system and needs a thermal shield to cut off thermal radiation from the components of room temperature. The vacuum vessel thermal shield (VVTS) cooled to 70 K is placed in the narrow gap between the 5 K TF magnets and the 300 K vacuum vessel (VV). The VVTS is designed to be divided into 16 assembly modules of 22.5 deg. sector, each unit has an electrical insulation along the center line in the toroidal direction and four insulations in the poloidal direction to reduce eddy currents induced during plasma operations. All connections are bolted. The VVTS becomes consequently a rigid torus composed of 64 electrically insulated pieces. A key point of designing the VVTS is that supports of the VVTS are to be flexible enough to allow thermal constriction during cooling down to 70 K as well as sufficiently strong to withstand electromagnetic (EM) forces exerted on the VVTS during plasma disruptions. Leaf spring type supports devised to satisfy these requirements are to be installed along the mid plane of the VVTS. The cryopanel of the VVTS is of quilted plate type whose total thickness is 12 mm, cooled by 60 K, 20 bar GHe.

  15. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  16. Engineering development aspects of the HL-2A tokamak

    International Nuclear Information System (INIS)

    The HL-2A tokamak (design values: major radius 1.65 m, minor radius 0.4 m, plasma current 0.48 MA and toroidal field 2.8 T) is the first tokamak with an operating divertor in China. It is characterized by a large closed divertor chamber. This unique feature will make significant contributions to enhance our understanding of complex divertor plasma physics and to help validating divertor physics modelings. The engineering design, development, testing and commissioning of the HL-2A tokamak are described in this paper. Preliminary results show that the HL-2A tokamak has been successfully operated in the divertor configuration. The major parameters: plasma current Ip=168 kA, toroidal field BT=1.4 T, plasma line average density ne=1.7 x 1019 m-3, limiting vacuum pv=4.6 x 10-6 Pa, were achieved at the end of 2003. (authors)

  17. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report

  18. Results and Future Plan of JT-60U towards Steady-State Tokamak Reactor

    Institute of Scientific and Technical Information of China (English)

    S. Sakurai; the JT-60 Team

    2004-01-01

    Recent results of JT-60U towards establishment of physics basis for ITER and advanced tokamak operation are presented. Progress in high integrated performance is achieved with improvement of N-NB and ECRF heating systems. In the next experimental campaign 2003~2004, discharge duration with 17 MW heating will be extended up to 30 s for sustaining high-beta plasma longer than the current diffusion time. Superconducting modification of JT-60is planned to demonstrate high-beta plasma sustainment exceeding ideal MHD instability limit without wall stabilization.

  19. Study of the generation of non-inductive current in Tore Supra and application to operational scenarios of a continuous tokamak; Etude de la generation de courant non inductive dans Tore Supra et application aux scenarios operationnels d`un tokamak continu

    Energy Technology Data Exchange (ETDEWEB)

    Kazarian-Vibert, F.

    1996-07-05

    Lower Hybrid Current Drive in tokamak plasma allows to obtain continuous operations, which constitute a necessary step towards a definition of a thermonuclear fusion reactor. The objectives of this work is to define and study fully non inductive steady-state scenarios on Tore Supra. The current diffusion equation is solved to determined precisely the inductive and non inductive current density profiles and their influence on the time evolution of a discharge. Then, a new operation mode is studied theoretically and experimentally. In this scenario, the transformer primary circuit voltage is controlled in such a way that the flux consumption vanishes. It allows to achieve full steady-state discharges in a fast and reproducible manner. A theoretical flux consumption scaling law during plasma current ramp-up assisted by Lower Hybrid waves is presented and validated by experimental data, in view to minimized this consumption. The influence of a non monotonic current density profile on the confinement and the transport of energy in the plasma is also clearly illustrated by experiments. (author). 155 refs.

  20. Study of the non inductive current generation in Tore Supra and application to the operational scenario of a continuous tokamak; Etude de la generation de courant non inductive dans Tore Supra et application aux scenarios operationnels d`un tokamak continu

    Energy Technology Data Exchange (ETDEWEB)

    Kazarian-Vibert, F.

    1996-07-05

    Lower Hybrid Current Drive in tokamak plasmas allows to obtain continuous operations, which constitute a necessary step towards a definition of a thermonuclear fusion reactor. The objectives of this work is to define and study fully non inductive steady-state scenarios on Tore Supra. The current diffusion equation is solved to determined precisely the inductive and non inductive current density profiles and their influence on thee time evolution of a discharge. Then, a new operation mode is studied theoretically and experimentally. In this scenario, the transformer primary circuit voltage is controlled in such a way that the flux consumption vanishes. It allows to achieve full steady-state discharges in a fast and reproducible manner. A theoretical flux consumption scaling law during plasma current ramp-up assisted by Lower-Hybrid waves is presented and validated by experimental data, in view to minimized this consumption. The influence of a non monotonic current profile on the confinement and the transport of energy in the plasma is also clearly illustrated by experiments. (author). 138 refs., 16 figs., 1 tab.

  1. Innovative Bimolecular-Based Advanced Logic Operations: A Prime Discriminator and An Odd Parity Checker.

    Science.gov (United States)

    Zhou, Chunyang; Liu, Dali; Dong, Shaojun

    2016-08-17

    Herein, a novel logic operation of prime discriminator is first performed for the function of identifying the prime numbers from natural numbers less than 10. The prime discriminator logic operation is developed by DNA hybridizations and the conjugation of graphene oxide and single-stranded DNA as a reacting platform. On the basis of the similar reaction principle, an odd parity checker is also developed. The odd parity checker logic operation can identify the even numbers and odd numbers from natural numbers less than 10. Such advanced logic operations with digital recognition ability can provide a new field of vision toward prototypical DNA-based logic operations and promote the development of advanced logic circuits. PMID:27459592

  2. SUPPRESSION OF TEARING MODES BY MEANS OF LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    The onset of tearing modes and the resulting negative effects on plasma performance set significant limits on the operational domain of tokamaks. Modes with toroidal mode number (n) larger than two cause only a minor reduction in energy confinement (<10%). Modes which have a dominant poloidal mode number (m) of three and n=2 lead to a significant reduction in confinement (<30%) at fixed power. The plasma pressure β (normalized to the magnetic field pressure) can be raised further, albeit with very small incremental confinement. Pushing to higher β often destabilizes the m=2/n=1 tearing mode which can lock to the wall and lead to a complete and rapid disruption of the plasma with potentially serious consequences for the tokamak. The β values at which these modes usually appear in conventional tokamak discharges are well below the limits calculated using ideal MHD theory. Therefore, the tearing modes can set effective upper limits on energy confinement and pressure. Significant progress has been made in stabilizing these modes by local current generation using electron cyclotron waves. The tearing mode is essentially a deficit in current flowing helically, resonant with the spatial structure of the local magnetic field. This forms an ''island'' where the magnetic flux is no longer monotonic. It was predicted theoretically [1,2] that replacement of this ''missing'' current would return the plasma to the state prior to the instability. Experiments on the ASDEX-Upgrade [3], JT-60U [4], and DIII-D [5] tokamaks have demonstrated stabilization of m=3/n=2 modes using electron cyclotron current drive (ECCD) to replace the current in the island. Following these initial experiments, recent work on the DIII-D tokamak has demonstrated two significant advances in application of this technique--extending the operational domain stable to m=3/n=2 modes to higher β and the first suppression of the more dangerous m=2/n=1 mode

  3. Effects of an Advanced Reactor’s Design, Use of Automation, and Mission on Human Operators

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey C. Joe; Johanna H. Oxstrand

    2014-06-01

    The roles, functions, and tasks of the human operator in existing light water nuclear power plants (NPPs) are based on sound nuclear and human factors engineering (HFE) principles, are well defined by the plant’s conduct of operations, and have been validated by years of operating experience. However, advanced NPPs whose engineering designs differ from existing light-water reactors (LWRs) will impose changes on the roles, functions, and tasks of the human operators. The plans to increase the use of automation, reduce staffing levels, and add to the mission of these advanced NPPs will also affect the operator’s roles, functions, and tasks. We assert that these factors, which do not appear to have received a lot of attention by the design engineers of advanced NPPs relative to the attention given to conceptual design of these reactors, can have significant risk implications for the operators and overall plant safety if not mitigated appropriately. This paper presents a high-level analysis of a specific advanced NPP and how its engineered design, its plan to use greater levels of automation, and its expanded mission have risk significant implications on operator performance and overall plant safety.

  4. SST and ADITYA tokamak research in India

    International Nuclear Information System (INIS)

    Steady state operation of tokamaks plays an important role in high temperature magnetically confined plasma research. Steady state Superconducting Tokamak (SST) programme in India deals with the development of various technologies in this direction. SST-1 machine has been engineered and is being fabricated at the Institute for Plasma Research. The objectives of the machine are to study physics of plasma processes under steady state condition and develop the technologies related to steady state operation. Various sub-systems are being prototyped and developed. SST-1 is a large aspect ratio machine with a major radius of 1.1 m and a plasma minor radius of 0.2 m with elongation of 1.7 to 1.9 and triangularity of 0.5 to 0.7. It has been designed for 1000 sec operation at 3 T toroidal magnetic eld. Neutral beam Injection and Radio frequency heating systems are being developed to heat the plasma. Lower hybrid Current Drive system would sustain 200 kA of plasma current during 1000 sec operation. ADITYA tokamak has been upgraded with new diagnostics and RF heating systems. Thomson Scattering and ECE diagnostics have been operated. 200 kW Ion Cyclotron Resonance Heating (ICRH) and 200 kW Electron Cyclotron Resonance Heating (ECRH) systems have been successfully commissioned. RF assisted initial breakdown experiments have been initiated with these systems. (author)

  5. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  6. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  7. Research using small tokamaks

    International Nuclear Information System (INIS)

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  8. Research using small tokamaks

    International Nuclear Information System (INIS)

    This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs

  9. Free-boundary simulations of ITER advanced scenarios

    International Nuclear Information System (INIS)

    The successful operation of ITER advanced scenarios is likely to be a major step forward in the development of controlled fusion as a power production source. ITER advanced scenarios raise specific challenges that are not encountered in presently-operated tokamaks. In this thesis, it is argued that ITER advanced operation may benefit from optimal control techniques. Optimal control ensures high performance operation while guaranteeing tokamak integrity. The application of optimal control techniques for ITER operation is assessed and it is concluded that robust optimisation is appropriate for ITER operation of advanced scenarios. Real-time optimisation schemes are discussed and it is concluded that the necessary conditions of optimality tracking approach may potentially be appropriate for ITER operation, thus offering a viable closed-loop optimal control approach. Simulations of ITER advanced operation are necessary in order to assess the present ITER design and uncover the main difficulties that may be encountered during advanced operation. The DINA-CH and CRONOS full tokamak simulator is used to simulate the operation of the ITER hybrid and steady-state scenarios. It is concluded that the present ITER design is appropriate for performing a hybrid scenario pulse lasting more than 1000 sec, with a flat-top plasma current of 12 MA, and a fusion gain of Q ≅ 8. Similarly, a steady-state scenario without internal transport barrier, with a flat-top plasma current of 10 MA, and with a fusion gain of Q ≅ 5 can be realised using the present ITER design. The sensitivity of the advanced scenarios with respect to transport models and physical assumption is assessed using CRONOS. It is concluded that the hybrid scenario and the steady-state scenario are highly sensitive to the L-H transition timing, to the value of the confinement enhancement factor, to the heating and current drive scenario during ramp-up, and, to a lesser extent, to the density peaking and pedestal

  10. Sensitivity analysis of scenario models for operational risk Advanced Measurement Approach

    OpenAIRE

    Chaudhary, Dinesh

    2014-01-01

    Scenario Analysis (SA) plays a key role in determination of operational risk capital under Basel II Advanced Measurement Approach. However, operational risk capital based on scenario data may exhibit high sensitivity or wrong-way sensitivity to scenario inputs. In this paper, we first discuss scenario generation using quantile approach and parameter estimation using quantile matching. Then we use single-loss approximation (SLA) to examine sensitivity of scenario based capital to scenario inputs.

  11. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Energy Technology Data Exchange (ETDEWEB)

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  12. Incentives to strengthen international co-operation in R and D for advanced nuclear power technology

    International Nuclear Information System (INIS)

    This paper is concerned with the need for International Co-operation in R and D for Advanced Reactors in order to maintain options for the future deployment of nuclear power against the current background of declining R and D capability in Europe

  13. Software Systems 2--Compiler and Operating Systems Lab--Advanced, Data Processing Technology: 8025.33.

    Science.gov (United States)

    Dade County Public Schools, Miami, FL.

    The course outline has been prepared as a guide to help the student develop the skills and knowledge necessary to succeed in the field of data processing. By learning the purpose and principles of compiler programs and operating systems, the student will become familiar with advanced data processing procedures that are representative of computer…

  14. Continued advancement of the programming language HAL to an operational status

    Science.gov (United States)

    1971-01-01

    The continued advancement of the programming language HAL to operational status is reported. It is demonstrated that the compiler itself can be written in HAL. A HAL-in-HAL experiment proves conclusively that HAL can be used successfully as a compiler implementation tool.

  15. Advanced light source's approach to ensure conditions for safe top-off operation

    International Nuclear Information System (INIS)

    The purpose of this document is to outline the Advanced Light Source (ALS) approach for preventing a radiation accident scenario on the ALS experimental floor due to top-off operation. The document will describe the potential risks, the analysis, and the resulting specifications for the controls.

  16. DEVELOPMENT OF OPERATIONAL CONCEPTS FOR ADVANCED SMRs: THE ROLE OF COGNITIVE SYSTEMS ENGINEERING

    Energy Technology Data Exchange (ETDEWEB)

    Jacques Hugo; David Gertman

    2014-04-01

    Advanced small modular reactors (AdvSMRs) will use advanced digital instrumentation and control systems, and make greater use of automation. These advances not only pose technical and operational challenges, but will inevitably have an effect on the operating and maintenance (O&M) cost of new plants. However, there is much uncertainty about the impact of AdvSMR designs on operational and human factors considerations, such as workload, situation awareness, human reliability, staffing levels, and the appropriate allocation of functions between the crew and various automated plant systems. Existing human factors and systems engineering design standards and methodologies are not current in terms of human interaction requirements for dynamic automated systems and are no longer suitable for the analysis of evolving operational concepts. New models and guidance for operational concepts for complex socio-technical systems need to adopt a state-of-the-art approach such as Cognitive Systems Engineering (CSE) that gives due consideration to the role of personnel. This approach we report on helps to identify and evaluate human challenges related to non-traditional concepts of operations. A framework - defining operational strategies was developed based on the operational analysis of Argonne National Laboratory’s Experimental Breeder Reactor-II (EBR-II), a small (20MWe) sodium-cooled reactor that was successfully operated for thirty years. Insights from the application of the systematic application of the methodology and its utility are reviewed and arguments for the formal adoption of CSE as a value-added part of the Systems Engineering process are presented.

  17. Preparing GMAT for Operational Maneuver Planning of the Advanced Composition Explorer (ACE)

    Science.gov (United States)

    Qureshi, Rizwan Hamid; Hughes, Steven P.

    2014-01-01

    The General Mission Analysis Tool (GMAT) is an open-source space mission design, analysis and trajectory optimization tool. GMAT is developed by a team of NASA, private industry, public and private contributors. GMAT is designed to model, optimize and estimate spacecraft trajectories in flight regimes ranging from low Earth orbit to lunar applications, interplanetary trajectories and other deep space missions. GMAT has also been flight qualified to support operational maneuver planning for the Advanced Composition Explorer (ACE) mission. ACE was launched in August, 1997 and is orbiting the Sun-Earth L1 libration point. The primary science objective of ACE is to study the composition of both the solar wind and the galactic cosmic rays. Operational orbit determination, maneuver operations and product generation for ACE are conducted by NASA Goddard Space Flight Center (GSFC) Flight Dynamics Facility (FDF). This paper discusses the entire engineering lifecycle and major operational certification milestones that GMAT successfully completed to obtain operational certification for the ACE mission. Operational certification milestones such as gathering of the requirements for ACE operational maneuver planning, gap analysis, test plans and procedures development, system design, pre-shadow operations, training to FDF ACE maneuver planners, shadow operations, Test Readiness Review (TRR) and finally Operational Readiness Review (ORR) are discussed. These efforts have demonstrated that GMAT is flight quality software ready to support ACE mission operations in the FDF.

  18. A novel approach to linearization of the electromagnetic parameters of tokamaks with an iron core

    International Nuclear Information System (INIS)

    The equivalent model of an iron core tokamak is developed, in which the electromagnetic parameters of several pairs of coils in opposite series (PCOS) are not dependent on the saturation of the iron core during tokamak operation. With this the electromagnetic parameters of all the coils in an iron core tokamak can be linearized, As an example, the electromagnetic parameters of Hefei Super-conductive Tokamak with iron core (HT-7) are linearized, and it is in good agreement with the experimental results. The linearization approach can be applied in real time plasma control and electromagnetic analysis

  19. A novel approach to linearization of the electromagnetic parameters of tokamaks with an iron core

    Energy Technology Data Exchange (ETDEWEB)

    Fu, P. E-mail: fupeng@mail.ipp.ac.cn; Liu, Z.Z.; Zou, J.H

    2002-05-01

    The equivalent model of an iron core tokamak is developed, in which the electromagnetic parameters of several pairs of coils in opposite series (PCOS) are not dependent on the saturation of the iron core during tokamak operation. With this the electromagnetic parameters of all the coils in an iron core tokamak can be linearized, As an example, the electromagnetic parameters of Hefei Super-conductive Tokamak with iron core (HT-7) are linearized, and it is in good agreement with the experimental results. The linearization approach can be applied in real time plasma control and electromagnetic analysis.

  20. Draft Function Allocation Framework and Preliminary Technical Basis for Advanced SMR Concepts of Operations

    Energy Technology Data Exchange (ETDEWEB)

    Jacques Hugo; John Forester; David Gertman; Jeffrey Joe; Heather Medema; Julius Persensky; April Whaley

    2013-08-01

    This report presents preliminary research results from the investigation into the development of new models and guidance for Concepts of Operations in advanced small modular reactor (AdvSMR) designs. AdvSMRs are nuclear power plants (NPPs), but unlike conventional large NPPs that are constructed on site, AdvSMRs systems and components will be fabricated in a factory and then assembled on site. AdvSMRs will also use advanced digital instrumentation and control systems, and make greater use of automation. Some AdvSMR designs also propose to be operated in a multi-unit configuration with a single central control room as a way to be more cost-competitive with existing NPPs. These differences from conventional NPPs not only pose technical and operational challenges, but they will undoubtedly also have regulatory compliance implications, especially with respect to staffing requirements and safety standards.

  1. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Science.gov (United States)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  2. Next tokamak facility

    International Nuclear Information System (INIS)

    Design studies on a superconducting, long-pulse, current-driven, ignited tokamak, called the Toroidal Fusion Core Demonstration (TFCD), are being conducted by the Fusion Engineering Design Center (FEDC) and Princeton Plasma Physics Laboratory (PPPL) with additional broad community involvement. Options include the use of all-superconducting toroidal field (TF) coils, a superconducting-copper hybrid arrangement of TF coils, or all-copper TF coils. Only the first two options have been considered to date. The general feasibility of these approaches has been established with the goal of high performance (ignition, approx. 390 MW; wall loading approx. 2.2 MW/m2) at minimum capital cost. The preconceptual effort will be completed in early FY 1984 and a selection made from the indicated options. The TFCD is judged to represent a reasonable necessary step between the Tokamak Fusion Test Reactor (TFTR) and the Engineering Test Reactor

  3. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  4. Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

    International Nuclear Information System (INIS)

    The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs

  5. Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Higgins, J.; Nasta, K.

    1997-01-01

    The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs.

  6. [High beta tokamak research

    International Nuclear Information System (INIS)

    Our activities on High Beta Tokamak Research during the past 20 months of the present grant period can be divided into six areas: reconstruction and modeling of high beta equilibria in HBT; measurement and analysis of MHD instabilities observed in HBT; measurements of impurity transport; diagnostic development on HBT; numerical parameterization of the second stability regime; and conceptual design and assembly of HBT-EP. Each of these is described in some detail in the sections of this progress report

  7. Advanced, Integrated Control for Building Operations to Achieve 40% Energy Saving

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Yan; Song, Zhen; Loftness, Vivian; Ji, Kun; Zheng, Sam; Lasternas, Bertrand; Marion, Flore; Yuebin, Yu

    2012-10-15

    We developed and demonstrated a software based integrated advanced building control platform called Smart Energy Box (SEB), which can coordinate building subsystem controls, integrate variety of energy optimization algorithms and provide proactive and collaborative energy management and control for building operations using weather and occupancy information. The integrated control system is a low cost solution and also features: Scalable component based architecture allows to build a solution for different building control system configurations with needed components; Open Architecture with a central data repository for data exchange among runtime components; Extendible to accommodate variety of communication protocols. Optimal building control for central loads, distributed loads and onsite energy resource; uses web server as a loosely coupled way to engage both building operators and building occupants in collaboration for energy conservation. Based on the open platform of SEB, we have investigated and evaluated a variety of operation and energy saving control strategies on Carnegie Mellon University Intelligent Work place which is equipped with alternative cooling/heating/ventilation/lighting methods, including radiant mullions, radiant cooling/heating ceiling panels, cool waves, dedicated ventilation unit, motorized window and blinds, and external louvers. Based on the validation results of these control strategies, they were integrated in SEB in a collaborative and dynamic way. This advanced control system was programmed and computer tested with a model of the Intelligent Workplace's northern section (IWn). The advanced control program was then installed in the IWn control system; the performance was measured and compared with that of the state of the art control system to verify the overall energy savings great than 40%. In addition advanced human machine interfaces (HMI's) were developed to communicate both with building

  8. Ideal MHD Stability Prediction and Required Power for EAST Advanced Scenario

    Science.gov (United States)

    Chen, Junjie; Li, Guoqiang; Qian, Jinping; Liu, Zixi

    2012-11-01

    The Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. The ideal magnetohydrodynamic (MHD) stability and required power for the EAST advanced tokamak (AT) scenario with negative central shear and double transport barrier (DTB) are investigated. With the equilibrium code TOQ and stability code GATO, the ideal MHD stability is analyzed. It is shown that a moderate ratio of edge transport barriers' (ETB) height to internal transport barriers' (ITBs) height is beneficial to ideal MHD stability. The normalized beta βN limit is about 2.20 (without wall) and 3.70 (with ideal wall). With the scaling law of energy confinement time, the required heating power for EAST AT scenario is calculated. The total heating power Pt increases as the toroidal magnetic field BT or the normalized beta βN is increased.

  9. Ideal MHD Stability Prediction and Required Power for EAST Advanced Scenario

    Institute of Scientific and Technical Information of China (English)

    陈均杰; 李国强; 钱金平; 刘子奚

    2012-01-01

    The Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. The ideal magnetohydrodynamic (MHD) stability and required power for the EAST advanced tokamak (AT) scenario with negative central shear and double transport barrier (DTB) are investigated. With the equilibrium code TOQ and stability code GATO, the ideal MHD stability is analyzed. It is shown that a moderate ratio of edge transport barriers' (ETB) height to internal transport barriers' (ITBs) height is beneficial to ideal MHD stability. The normalized beta/3N limit is about 2.20 (without wall) and 3.70 (with ideal wall). With the scaling law of energy confinement time, the required heating power for EAST AT scenario is calculated. The total heating power Pt increases as the toroidal magnetic field BT or the normalized beta βN is increased.

  10. Operation flexibility and availability improvements using BEACON, an advanced core monitoring system

    International Nuclear Information System (INIS)

    In response to utilities needs in improving plant operation flexibility and plant availability, Westinghouse introduced the advanced core monitoring and operational support system, BEACON, two years ago. Since then, the continuous development of the BEACON system has led to significant advances in further reducing utilities Operation and Maintenance (O and M) costs. The development of the BEACON system is made possible by two breakthroughs: 1) advanced numerical method to solve the diffusion equations extremely fast and 2) development of cost effective, state-of-the-art computing system, workstation. This paper presents the numerical scheme used in the neutronic solution and how BEACON uses the core instrumentations to provide the continuous three-dimensional (3D) core power distribution. Once the state of the core is known on a continuous basis, several indirect surveillance and/or Technical Specifications on core power distribution can be relaxed or totally eliminated. Section 1 outlines the numerical scheme used in BEACON for solving the diffusion equations and to provide the 3D continuous power distribution. Section 2 describes the hardware requirements. Section 3 discusses applications of BEACON to improve plant operation flexibility and plant availability. Examples of actual BEACON usage to demonstrate its effectiveness are presented in Section 4 and the paper is closed with a summary of future directions. (author). 4 refs, 6 figs

  11. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    NARCIS (Netherlands)

    Box, F. M. A.; Howard, J.; VandeKolk, E.; Meijer, F. G.

    1997-01-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. S

  12. Real-time control of Tokamak plasmas: from control of physics to physics-based control

    International Nuclear Information System (INIS)

    Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a digital realtime control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This experimental platform has been used to develop and test new control strategies for three plasma physics instabilities: sawtooth, edge localized mode (ELM) and neoclassical tearing mode (NTM). The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface (q: safety factor). A sawtooth pacing controller was developed which is able to control the time of appearance of the next sawtooth crash. Each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes have been used to study the relationship between sawteeth and NTMs. Postcrash MHD activity can provide the ‘seed’ island for an NTM, which then grows under its neoclassical bootstrap drive. The seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have

  13. Polarimetric spectra analysis for tokamak pitch angle measurements

    International Nuclear Information System (INIS)

    Measurements of the internal magnetic field structures using conventional polarimetric approaches are considered extremely challenging in fusion-reactor environments whereas the information on current density profiles is essential to establish steady-state and advance operation scenarios in such reactor-relevant devices. Therefore, on ITER a hybrid system is proposed for the current density measurements that uses both polarimetry and spectral measurements. The spectrum-based approaches have been tested in the Korea Superconducting Tokamak Advanced Research (KSTAR) during the past two plasma campaigns. As such, KSTAR is a test-bed for the proposed ITER hybrid system. Measurements in the plasma core are based on the motional Stark effect (MSE) spectrum of the neutral beam emission. For the edge profiles, the Zeeman effect (ZE) acting on the lithium emission spectrum of the newly installed (2013) Lithium-beam-diagnostic is exploited. The neutral beam emission spectra, complicated by the multi-ion-source beam injection, are successfully fitted making use of the data provided by the Atomic Data and Analysis Structure (ADAS) database package. This way pitch angle profiles could be retrieved from the beam emission spectra. With the same spectrometer/CCD hardware as on MSE, but with a different wavelength range and different lines of sight, the first ZE spectrum measurements have been made. The Zeeman splitting comparable to and greater than the instrumental broadening has been routinely detected at high toroidal field operations ( ∼ 3 Tesla)

  14. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  15. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  16. MHD analysis of edge instabilities in the JET tokamak

    NARCIS (Netherlands)

    Perez von Thun, Christian Pedro

    2004-01-01

    The aim of nuclear fusion energy research is to demonstrate the feasibility of nuclear fusion reactors as a future energy source. The tokamak is the most advanced fusion machine to date, and is most likely the first system to be converted into a reactor. An important subject of nuclear fusion resear

  17. Water chemistry control to meet the advanced design and operation of light water reactors

    International Nuclear Information System (INIS)

    Water chemistry control is one of the key technologies to establish safe and reliable operation of nuclear power plants. The road maps on R and D plans for water chemistry of nuclear power systems in Japan have been proposed along with promotion of R and D related water chemistry improvement for the advanced application of light water reactors (LWRs). The technical trends were divided into four categories, dose rate reduction, structural integrity, fuel integrity and radioactive waste reduction, and latest technical break through for each category was shown for the advanced application of LWRs. At the same time, the technical break through and the latest movements for regulation of water chemistry were introduced for each of major organizations related to nuclear engineering in the world. The conclusions were summarized as follows; 1. Water chemistry improvements might contribute to achieve the advanced application of LWRs, while water chemistry should be often changed to achieve the advanced application of LWRs. 2. Only one solution for water chemistry control was not obtained for achieving the advanced application of LWRs, but miscellaneous solutions were possible for achieving one. Optimal water chemistry control was desired for having the good practices for satisfying multi-targets at the same time and it was much affected by the plant unique systems and operational history. 3. That meant it was difficult to determine water chemistry regulation targets for achieving application of LWRs but it was necessary to prepare suitable guideline for good achievement of application of LWRs. That meant the guideline should be recommendation for good practice in the plant. 4. The water chemistry guide line should be modified along with progress of plant operation and water chemistry and related technologies. (author)

  18. Towards energy efficient operation of Heating, Ventilation and Air Conditioning systems via advanced supervisory control design

    Science.gov (United States)

    Oswiecinska, A.; Hibbs, J.; Zajic, I.; Burnham, K. J.

    2015-11-01

    This paper presents conceptual control solution for reliable and energy efficient operation of heating, ventilation and air conditioning (HVAC) systems used in large volume building applications, e.g. warehouse facilities or exhibition centres. Advanced two-level scalable control solution, designed to extend capabilities of the existing low-level control strategies via remote internet connection, is presented. The high-level, supervisory controller is based on Model Predictive Control (MPC) architecture, which is the state-of-the-art for indoor climate control systems. The innovative approach benefits from using passive heating and cooling control strategies for reducing the HVAC system operational costs, while ensuring that required environmental conditions are met.

  19. Modern advances to the Modular Fly-Away Kit (MFLAK) to support maritime interdiction operations

    OpenAIRE

    Cross, Eric C.

    2007-01-01

    This thesis will test the performance of an end-to-end network solution designed to augment Maritime Interdiction Operations that support boarding parties and their near real time communications with supporting agencies. The 802.16 point-to-point and point-to-multipoint Orthogonal Frequency Divisional Multiplexing (OFDM) shall be upgraded to reflect modern advances in 802.16. Additionally, there will be several enhancements to the peripherals associated with end user innovations and they will...

  20. Using Micro-Synchrophasor Data for Advanced Distribution Grid Planning and Operations Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, Emma [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Kiliccote, Sila [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); McParland, Charles [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Roberts, Ciaran [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2014-07-01

    This report reviews the potential for distribution-grid phase-angle data that will be available from new micro-synchrophasors (µPMUs) to be utilized in existing distribution-grid planning and operations analysis. This data could augment the current diagnostic capabilities of grid analysis software, used in both planning and operations for applications such as fault location, and provide data for more accurate modeling of the distribution system. µPMUs are new distribution-grid sensors that will advance measurement and diagnostic capabilities and provide improved visibility of the distribution grid, enabling analysis of the grid’s increasingly complex loads that include features such as large volumes of distributed generation. Large volumes of DG leads to concerns on continued reliable operation of the grid, due to changing power flow characteristics and active generation, with its own protection and control capabilities. Using µPMU data on change in voltage phase angle between two points in conjunction with new and existing distribution-grid planning and operational tools is expected to enable model validation, state estimation, fault location, and renewable resource/load characterization. Our findings include: data measurement is outstripping the processing capabilities of planning and operational tools; not every tool can visualize a voltage phase-angle measurement to the degree of accuracy measured by advanced sensors, and the degree of accuracy in measurement required for the distribution grid is not defined; solving methods cannot handle the high volumes of data generated by modern sensors, so new models and solving methods (such as graph trace analysis) are needed; standardization of sensor-data communications platforms in planning and applications tools would allow integration of different vendors’ sensors and advanced measurement devices. In addition, data from advanced sources such as µPMUs could be used to validate models to improve

  1. Simulation of EAST vertical displacement events by tokamak simulation code

    Science.gov (United States)

    Qiu, Qinglai; Xiao, Bingjia; Guo, Yong; Liu, Lei; Xing, Zhe; Humphreys, D. A.

    2016-10-01

    Vertical instability is a potentially serious hazard for elongated plasma. In this paper, the tokamak simulation code (TSC) is used to simulate vertical displacement events (VDE) on the experimental advanced superconducting tokamak (EAST). Key parameters from simulations, including plasma current, plasma shape and position, flux contours and magnetic measurements match experimental data well. The growth rates simulated by TSC are in good agreement with TokSys results. In addition to modeling the free drift, an EAST fast vertical control model enables TSC to simulate the course of VDE recovery. The trajectories of the plasma current center and control currents on internal coils (IC) fit experimental data well.

  2. Adaptive grid finite element model of the tokamak scrapeoff layer

    Energy Technology Data Exchange (ETDEWEB)

    Kuprat, A.P.; Glasser, A.H. [Los Alamos National Lab., NM (United States)

    1995-07-01

    The authors discuss unstructured grids for application to transport in the tokamak edge SOL. They have developed a new metric with which to judge element elongation and resolution requirements. Using this method, the authors apply a standard moving finite element technique to advance the SOL equations while inserting/deleting dynamically nodes that violate an elongation criterion. In a tokamak plasma, this method achieves a more uniform accuracy, and results in highly stretched triangular finite elements, except near separatrix X-point where transport is more isotropic.

  3. DIII-D research operations annual report to the U.S. Department of Energy, October 1, 1995--September 30, 1996

    International Nuclear Information System (INIS)

    The mission of the DIII-D research program is to advance fusion energy science understanding and predictive capability and to improve and optimize the tokamak concept. A long term goal remains to integrate these products into a demonstration of high confinement, high plasma pressure (plasma β), sustained long pulse operation with fusion power plant relevant heat and particle handling capability. The DIII-D program is a world recognized leader in tokamak concept improvement and a major contributor to the physics R and D needs of the International Thermonuclear Experimental Reactor (ITER). The scientific objectives of the DIII-D program are given in Table 1-2. The FY96 DIII-D research program was highly successful, as described in this report. A moderate sized tokamak, DIII-D is a world leader in tokamak innovation with exceptional performance, measured in normalized parameters

  4. DIII-D research operations annual report to the U.S. Department of Energy, October 1, 1995--September 30, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The mission of the DIII-D research program is to advance fusion energy science understanding and predictive capability and to improve and optimize the tokamak concept. A long term goal remains to integrate these products into a demonstration of high confinement, high plasma pressure (plasma {beta}), sustained long pulse operation with fusion power plant relevant heat and particle handling capability. The DIII-D program is a world recognized leader in tokamak concept improvement and a major contributor to the physics R and D needs of the International Thermonuclear Experimental Reactor (ITER). The scientific objectives of the DIII-D program are given in Table 1-2. The FY96 DIII-D research program was highly successful, as described in this report. A moderate sized tokamak, DIII-D is a world leader in tokamak innovation with exceptional performance, measured in normalized parameters.

  5. Data acquisition and processing system of the electron cyclotron emission imaging system of the KSTAR tokamak

    International Nuclear Information System (INIS)

    A new innovative electron cyclotron emission imaging (ECEI) diagnostic system for the Korean Superconducting Tokamak Advanced Research (KSTAR) produces a large amount of data. The design of the data acquisition and processing system of the ECEI diagnostic system should consider covering the large data production and flow. The system design is based on the layered structure scalable to the future extension to accommodate increasing data demands. Software architecture that allows a web-based monitoring of the operation status, remote experiment, and data analysis is discussed. The operating software will help machine operators and users validate the acquired data promptly, prepare next discharge, and enhance the experiment performance and data analysis in a distributed environment.

  6. Advancing Data Assimilation in Operational Hydrologic Forecasting: Progresses, Challenges, and Emerging Opportunities

    Science.gov (United States)

    Liu, Yuqiong; Weerts, A.; Clark, M.; Hendricks Franssen, H.-J; Kumar, S.; Moradkhani, H.; Seo, D.-J.; Schwanenberg, D.; Smith, P.; van Dijk, A. I. J. M.; van Velzen, N.; He, M.; Lee, H.; Noh, S. J.; Rakovec, O.; Restrepo, P.

    2012-01-01

    Data assimilation (DA) holds considerable potential for improving hydrologic predictions as demonstrated in numerous research studies. However, advances in hydrologic DA research have not been adequately or timely implemented in operational forecast systems to improve the skill of forecasts for better informed real-world decision making. This is due in part to a lack of mechanisms to properly quantify the uncertainty in observations and forecast models in real-time forecasting situations and to conduct the merging of data and models in a way that is adequately efficient and transparent to operational forecasters. The need for effective DA of useful hydrologic data into the forecast process has become increasingly recognized in recent years. This motivated a hydrologic DA workshop in Delft, the Netherlands in November 2010, which focused on advancing DA in operational hydrologic forecasting and water resources management. As an outcome of the workshop, this paper reviews, in relevant detail, the current status of DA applications in both hydrologic research and operational practices, and discusses the existing or potential hurdles and challenges in transitioning hydrologic DA research into cost-effective operational forecasting tools, as well as the potential pathways and newly emerging opportunities for overcoming these challenges. Several related aspects are discussed, including (1) theoretical or mathematical aspects in DA algorithms, (2) the estimation of different types of uncertainty, (3) new observations and their objective use in hydrologic DA, (4) the use of DA for real-time control of water resources systems, and (5) the development of community-based, generic DA tools for hydrologic applications. It is recommended that cost-effective transition of hydrologic DA from research to operations should be helped by developing community-based, generic modeling and DA tools or frameworks, and through fostering collaborative efforts among hydrologic modellers, DA

  7. Consensus statement on advancing research in emergency department operations and its impact on patient care.

    Science.gov (United States)

    Yiadom, Maame Yaa A B; Ward, Michael J; Chang, Anna Marie; Pines, Jesse M; Jouriles, Nick; Yealy, Donald M

    2015-06-01

    The consensus conference on "Advancing Research in Emergency Department (ED) Operations and Its Impact on Patient Care," hosted by The ED Operations Study Group (EDOSG), convened to craft a framework for future investigations in this important but understudied area. The EDOSG is a research consortium dedicated to promoting evidence-based clinical practice in emergency medicine. The consensus process format was a modified version of the NIH Model for Consensus Conference Development. Recommendations provide an action plan for how to improve ED operations study design, create a facilitating research environment, identify data measures of value for process and outcomes research, and disseminate new knowledge in this area. Specifically, we call for eight key initiatives: 1) the development of universal measures for ED patient care processes; 2) attention to patient outcomes, in addition to process efficiency and best practice compliance; 3) the promotion of multisite clinical operations studies to create more generalizable knowledge; 4) encouraging the use of mixed methods to understand the social community and human behavior factors that influence ED operations; 5) the creation of robust ED operations research registries to drive stronger evidence-based research; 6) prioritizing key clinical questions with the input of patients, clinicians, medical leadership, emergency medicine organizations, payers, and other government stakeholders; 7) more consistently defining the functional components of the ED care system, including observation units, fast tracks, waiting rooms, laboratories, and radiology subunits; and 8) maximizing multidisciplinary knowledge dissemination via emergency medicine, public health, general medicine, operations research, and nontraditional publications. PMID:26014365

  8. Mechanical and thermal characteristics of JT-60 tokamak machine demonstrated in its power tests

    International Nuclear Information System (INIS)

    JT-60 power tests were carried out from Dec. 10, 1984 to Feb. 20, 1985 to demonstrate, in advance of actual plasma operation, satisfactory performance of tokamak machine, power suppliers and control system in combination. The tests began with low power test of individual coil systems and progressed to full power tests. Power tests were successfully concluded with the following conclusions. (1) All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and structural integrity of tokamak machine. (2) Measured strain and deflection showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. (3) Vibration of lateral port was found to be large up to 50 m/s2 and caused excessive vibration of gate-valves. (4) A few limitations to machine operation were made clear quantatively. (5) It was found that the existing detectors were insufficient to monitor the machine integrity and a few kinds of detectors were necessary to be installed. (author)

  9. Design and realization of central control system for J-TEXT tokamak

    International Nuclear Information System (INIS)

    To keep the J-TEXT tokamak running in order, it needs a central control system to coordinate the operations of all the sub-systems. The J-TEXT tokamak central control system adopts industrial PC with some PCI control cards based on Windows and QNX operation systems. This control system has been finished the test and put into use in the first-round experiment of the J-TEXT tokamak successfully. The testing and operation show that this system is stable and reliable. (authors)

  10. Engineering overview of the National Spherical Tokamak Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chrzanowski, J.H.; Fan, H.M.; Heitzenroeder, P.J.; Ono, M.; Robinson, J. [Princeton Plasma Physics Lab., NJ (United States)

    1995-12-31

    The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device designed for a plasma current of 1 MA. It features auxiliary heating and current drive and a close-fitting conducting shell to maximize plasma pressure. NSTX is designed for a 5 sec. experimental pulse to demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The device will be sited in the former Princeton Large Torus (PLT) test cell and will utilize the PLT radiation shielding, base structure, and cell utilities. NSTX will utilize the S-1 Spheromak vacuum vessel, poloidal field coils, and capacitor banks (for helicity injection). The poloidal Beta Experiment-Modified (PBX-M) power supplies will be shared to power the PF and TF coil systems. Existing RF hardware and infrastructure will be used for heating systems. TFTR data acquisition and diagnostics resources are planned to be used. In total, NSTX will utilize site credits with a value of {approximately}$50 M, reducing base construction cost of the device to $18.6 M. Twelve water-cooled copper demountable toroidal field (TF) coils provide the 5.4 kg (pulsed) and 3.5 kg (long pulse > 5 sec) toroidal field at the plasma center. Poloidal fields are generated by windings contained in the center column and four pairs water-cooled copper coils supported directly on the vacuum vessel. One of the most critical components of the device is the center stack, which consists of the inner legs of the TF coils overwrapped with ohmic heating and poloidal field windings. The ohmic heating coil windings are designed to optimize the V-s and together with the PF coils, produce a flux swing of 1 V-s.

  11. Tokamak burn control

    International Nuclear Information System (INIS)

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  12. Non-operative advances: what has happened in the last 50 years in paediatric surgery?

    Science.gov (United States)

    Holland, Andrew J A; McBride, Craig A

    2015-01-01

    Paediatric surgeons remain paediatric clinicians who have the unique skill set to treat children with surgical problems that may require operative intervention. Many of the advances in paediatric surgical care have occurred outside the operating theatre and have involved significant input from medical, nursing and allied health colleagues. The establishment of neonatal intensive care units, especially those focusing on the care of surgical infants, has greatly enhanced the survival rates and long-term outcomes of those infants with major congenital anomalies requiring surgical repair. Educational initiatives such as the advanced trauma life support and emergency management of severe burns courses have facilitated improved understanding and clinical care. Paediatric surgeons have led with the non-operative management of solid organ injury following blunt abdominal trauma. Nano-crystalline burn wound dressings have enabled a reduced frequency of painful dressing changes in addition to effective antimicrobial efficacy and enhanced burn wound healing. Burns care has evolved so that many children may now be treated almost exclusively in an ambulatory care setting or as day case-only patients, with novel technologies allowing accurate prediction of burn would outcome and planning of elective operative intervention to achieve burn wound closure. PMID:25588791

  13. ECH on the MTX [Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    The Microwave Tokamak Experiment (MTX) at LLNL is investigating the heating of high density Tokamak plasmas using an intense pulse FEL. Our first experiments, now beginning, will study the absorption and plasma heating of single FEL pulses (20 ns pulse length and peak power up to 2 GW) at a frequency of 140 GHz. A later phase of experiments also at 140 GHz will study FEL heating at 5 kHz rate for a pulse train up to 50 pulses (35 ns pulse length and peak power up to 4 GW). Future operations are planned at 250 GHz with an average power of 2 MW for a pulse train of 0.5 s. The microwave output of the FEL is transported quasi-optically to the tokamak through a window-less, evacuated pipe of 20 in. diameter, using a six mirror system. Computational modelling of the non-linear absorption for the MTX geometry predicts single-pass absorption of 40% at a density and temperature of 1.8 /times/ 1020m/sup /minus/3/ and 1 keV, respectively. To measure plasma microwave absorption and backscatter, diagnostics are available to measure forward and reflected power (parallel wire grid beam-splitter and mirror directional couplers) and power transmitted through the plasma (segmented calorimeter and waveguide detector). Other fast diagnostics include ECE, Thompson scattering, soft x-rays, and fast magnetic probes. 8 refs., 2 figs

  14. Intra-operative maternal complications of emergency cesarean section done in advanced labor

    International Nuclear Information System (INIS)

    Background: Emergency cesarean section done in advanced labor is a big challenge in obstetrics due to increased risk of intraoperative complications. In the last decade, a rapid increase in cesarean section done in advanced labor has been observed. Difficult deli-very of the fetal head during cesarean section carries a high risk of intraoperative complications like cervical and uterine tears, intra operative hemorrhage and trauma to the baby. Objectives: The purpose of this study is to find out the frequency and risk factors for intra-operative complications in emergency cesarean section done in advanced labor, so that appropriate management protocols can be planned to reduce these complications. Study Design: Prospective cohort study. Materials and Methods: This prospective study was carried out in Obstetrics and Gynecology Unit - 2 of Services Institute of Medical Sciences, Services Hospital, Lahore; from 1st January 2007 to 31st December 2007. All patients undergoing emergency cesarean sections done on laboring mothers were included in the study. The sample was divided into two groups; emergency C-section done in advanced labor as the study group and emergency C-section in early labor as the control group. Data were collected regarding age, parity, booked or unbooked status, indications for cesarean section, level of competence of operating surgeon, intra-operative complications and the risk factors for these complications. Data were recorded on a structured proforma and compared between the two groups. Statistical Analysis: Data were analyzed using computer programme SPSS Version 14 for windows applying student t-test for quantitative and chai square test for qualitative parameters. A p-value < 0.05 was used as statistically significant. Results: Out of 2064 total deliveries in the year 2007, 1290 (62.5%) were vaginal deliveries and 774 (37.5%) were C-Sections. Out of 774 C-Section, 174 (23%) were elective and 600 (77%) were emergency. Out of 600 emergency C

  15. Work Domain Analysis Methodology for Development of Operational Concepts for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This report describes a methodology to conduct a Work Domain Analysis in preparation for the development of operational concepts for new plants. This method has been adapted from the classical method described in the literature in order to better deal with the uncertainty and incomplete information typical of first-of-a-kind designs. The report outlines the strategy for undertaking a Work Domain Analysis of a new nuclear power plant and the methods to be used in the development of the various phases of the analysis. Basic principles are described to the extent necessary to explain why and how the classical method was adapted to make it suitable as a tool for the preparation of operational concepts for a new nuclear power plant. Practical examples are provided of the systematic application of the method and the various presentation formats in the operational analysis of advanced reactors.

  16. Advanced data management for optimising the operation of a full-scale WWTP.

    Science.gov (United States)

    Beltrán, Sergio; Maiza, Mikel; de la Sota, Alejandro; Villanueva, José María; Ayesa, Eduardo

    2012-01-01

    The lack of appropriate data management tools is presently a limiting factor for a broader implementation and a more efficient use of sensors and analysers, monitoring systems and process controllers in wastewater treatment plants (WWTPs). This paper presents a technical solution for advanced data management of a full-scale WWTP. The solution is based on an efficient and intelligent use of the plant data by a standard centralisation of the heterogeneous data acquired from different sources, effective data processing to extract adequate information, and a straightforward connection to other emerging tools focused on the operational optimisation of the plant such as advanced monitoring and control or dynamic simulators. A pilot study of the advanced data manager tool was designed and implemented in the Galindo-Bilbao WWTP. The results of the pilot study showed its potential for agile and intelligent plant data management by generating new enriched information combining data from different plant sources, facilitating the connection of operational support systems, and developing automatic plots and trends of simulated results and actual data for plant performance and diagnosis.

  17. Static and Dynamic Mechanical Analyses for the Vacuum Vessel of EAST Superconducting Tokamak Device

    Science.gov (United States)

    Song, Yuntao; Yao, Damao; Du, Shijun; Wu, Songtao; Weng, Peide

    2006-03-01

    EAST (experimental advanced superconducting tokamak) is an advanced steady-state plasma physics experimental device, which is being constructed as the Chinese National Nuclear Fusion Research Project. During the plasma operation the vacuum vessel as one of the key component will withstand the electromagnetic force due to the plasma disruption, the Halo current and the toroidal field coil quench, the pressure of boride water and the thermal load due to 250 oC baking by pressurized nitrogen gas. In this paper a report of the static and dynamic mechanical analyses of the vacuum vessel is made. Firstly the applied loads on the vacuum vessel were given and the static stress distribution under the gravitational loads, the pressure loads, the electromagnetic loads and thermal loads were investigated. Then a series of primary dynamic, buckling and fatigue life analyses were performed to predict the structure's dynamic behavior. A seismic analysis was also conducted.

  18. ITER tokamak device

    Science.gov (United States)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER, a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fueling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  19. Edge turbulence in tokamaks

    Science.gov (United States)

    Nedospasov, A. V.

    1992-12-01

    Edge turbulence is of decisive importance for the distribution of particle and energy fluxes to the walls of tokamaks. Despite the availability of extensive experimental data on the turbulence properties, its nature still remains a subject for discussion. This paper contains a review of the most recent theoretical and experimental studies in the field, including mainly the studies to which Wootton (A.J. Wooton, J. Nucl. Mater. 176 & 177 (1990) 77) referred to most in his review at PSI-9 and those published later. The available theoretical models of edge turbulence with volume dissipation due to collisions fail to fully interpret the entire combination of experimental facts. In the scrape-off layer of a tokamak the dissipation prevails due to the flow of current through potential shifts near the surface of limiters of divertor plates. The different origins of turbulence at the edge and in the core plasma due to such dissipation are discussed in this paper. Recent data on the electron temperature fluctuations enabled one to evaluate the electric probe measurements of turbulent flows of particles and heat critically. The latest data on the suppression of turbulence in the case of L-H transitions are given. In doing so, the possibility of exciting current instabilities in biasing experiments (rather than only to the suppression of existing turbulence) is given some attention. Possible objectives of further studies are also discussed.

  20. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  1. Deposit of thin films for Tokamaks conditioning

    International Nuclear Information System (INIS)

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (-6 to 4.5 x 10-6 Ω-m, thus taking the Zef value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow

  2. Real Time Hybrid Model Predictive Control for the Current Profile of the Tokamak à Configuration Variable (TCV

    Directory of Open Access Journals (Sweden)

    Izaskun Garrido

    2016-08-01

    Full Text Available Plasma stability is one of the obstacles in the path to the successful operation of fusion devices. Numerical control-oriented codes as it is the case of the widely accepted RZIp may be used within Tokamak simulations. The novelty of this article relies in the hierarchical development of a dynamic control loop. It is based on a current profile Model Predictive Control (MPC algorithm within a multiloop structure, where a MPC is developed at each step so as to improve the Proportional Integral Derivative (PID global scheme. The inner control loop is composed of a PID-based controller that acts over the Multiple Input Multiple Output (MIMO system resulting from the RZIp plasma model of the Tokamak à Configuration Variable (TCV. The coefficients of this PID controller are initially tuned using an eigenmode reduction over the passive structure model. The control action corresponding to the state of interest is then optimized in the outer MPC loop. For the sake of comparison, both the traditionally used PID global controller as well as the multiloop enhanced MPC are applied to the same TCV shot. The results show that the proposed control algorithm presents a superior performance over the conventional PID algorithm in terms of convergence. Furthermore, this enhanced MPC algorithm contributes to extend the discharge length and to overcome the limited power availability restrictions that hinder the performance of advanced tokamaks.

  3. The influence on the performance of operators along with the introduction of the advanced main control board

    International Nuclear Information System (INIS)

    This paper describes an influence on the performance of operators along with the introduction of the advanced main control board (MCB). The influence on the performance of operators is considered based on the operating procedure, the requirements for operators and the operator training. The operating procedure is the document which puts forward the way that the designer has thought in advance for the operators and describes the performance of operators. The introduction of the advanced MCB seems to be bringing a change of the operating procedure. The requirements for operators are the knowledge, skills and attitude, and crew resource management (CRM) skill. CRM skill makes use of the knowledge, skills and attitude and improves the team performance. The advanced MCB seems to induce a change of CRM skill i.e. the communication, decision making or problem solving, team building, situation awareness, and workload management of different shift teams. The operator training is the best way to verify the change of the operating procedure and CRM skill. (author)

  4. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Box, F.M.A.; Kolk, E. van de [Associatie Euratom-FOM, Nieuwegein (Netherlands). FOM-Instituut voor Plasmafysica; Howard, J. [Plasma Research Laboratory, Research School of Physical Science and Engineering, Australian National University, Canberra 0200 (Australia); Meijer, F.G. [Physics Faculty, University of Amsterdam, Amsterdam (Netherlands)

    1997-03-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. Several spectrometers, equipped with a charge-coupled device array, are being used with spectral ranges in the visible, the vacuum UV and the extreme UV. (orig.)

  5. Basic Physics of Tokamak Transport Final Technical Report.

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Amiya K.

    2014-05-12

    The goal of this grant has been to study the basic physics of various sources of anomalous transport in tokamaks. Anomalous transport in tokamaks continues to be one of the major problems in magnetic fusion research. As a tokamak is not a physics device by design, direct experimental observation and identification of the instabilities responsible for transport, as well as physics studies of the transport in tokamaks, have been difficult and of limited value. It is noted that direct experimental observation, identification and physics study of microinstabilities including ITG, ETG, and trapped electron/ion modes in tokamaks has been very difficult and nearly impossible. The primary reasons are co-existence of many instabilities, their broadband fluctuation spectra, lack of flexibility for parameter scans and absence of good local diagnostics. This has motivated us to study the suspected tokamak instabilities and their transport consequences in a simpler, steady state Columbia Linear Machine (CLM) with collisionless plasma and the flexibility of wide parameter variations. Earlier work as part of this grant was focused on both ITG turbulence, widely believed to be a primary source of ion thermal transport in tokamaks, and the effects of isotope scaling on transport levels. Prior work from our research team has produced and definitively identified both the slab and toroidal branches of this instability and determined the physics criteria for their existence. All the experimentally observed linear physics corroborate well with theoretical predictions. However, one of the large areas of research dealt with turbulent transport results that indicate some significant differences between our experimental results and most theoretical predictions. Latter years of this proposal were focused on anomalous electron transport with a special focus on ETG. There are several advanced tokamak scenarios with internal transport barriers (ITB), when the ion transport is reduced to

  6. Treatment of dairy manure using the microwave enhanced advanced oxidation process under a continuous mode operation.

    Science.gov (United States)

    Yu, Yang; Lo, Ing W; Liao, Ping H; Lo, Kwang V

    2010-11-01

    The microwave enhanced advanced oxidation process (MW/H(2)O(2)-AOP) was used to treat dairy manure for solubilization of nutrients and organic matters. This study investigated the effectiveness of the MW/H(2)O(2)-AOP under a continuous mode of operation, and compared the results to those of batch operations. The main factors affecting solubilization by the MW/H(2)O(2)-AOP were heating temperature and hydrogen peroxide dosage. Soluble chemical oxygen demand (SCOD) and volatile fatty acids (VFA) increased with an increase of microwave (MW) heating temperature; very high concentrations were obtained at 90°C. Insignificant amounts of ammonia and reducing sugars were released in all runs. An acidic pH condition was required for phosphorus solubilisation from dairy manure. The best yield was obtained at 90°C with an acid dosage of 1.0 %; about 92 % of total phosphorus and 90 % of total chemical oxygen demand were in the soluble forms. The MW/H(2)O(2)-AOP operated in a continuous operation mode showed pronounced synergistic effects between hydrogen peroxide and microwave irradiation when compared to a batch system under similar operating conditions, resulting in much better yields.

  7. Draft Function Allocation Framework and Preliminary Technical Basis for Advanced SMR Concepts of Operations

    Energy Technology Data Exchange (ETDEWEB)

    Jacques Hugo; John Forester; David Gertman; Jeffrey Joe; Heather Medema; Julius Persensky; April Whaley

    2013-04-01

    This report presents preliminary research results from the investigation in to the development of new models and guidance for concepts of operations (ConOps) in advanced small modular reactor (aSMR) designs. In support of this objective, three important research areas were included: operating principles of multi-modular plants, functional allocation models and strategies that would affect the development of new, non-traditional concept of operations, and the requiremetns for human performance, based upon work domain analysis and current regulatory requirements. As part of the approach for this report, we outline potential functions, including the theoretical and operational foundations for the development of a new functional allocation model and the identification of specific regulatory requirements that will influence the development of future concept of operations. The report also highlights changes in research strategy prompted by confirmationof the importance of applying the work domain analysis methodology to a reference aSMR design. It is described how this methodology will enrich the findings from this phase of the project in the subsequent phases and help in identification of metrics and focused studies for the determination of human performance criteria that can be used to support the design process.

  8. Operations management system advanced automation: Fault detection isolation and recovery prototyping

    Science.gov (United States)

    Hanson, Matt

    1990-01-01

    The purpose of this project is to address the global fault detection, isolation and recovery (FDIR) requirements for Operation's Management System (OMS) automation within the Space Station Freedom program. This shall be accomplished by developing a selected FDIR prototype for the Space Station Freedom distributed processing systems. The prototype shall be based on advanced automation methodologies in addition to traditional software methods to meet the requirements for automation. A secondary objective is to expand the scope of the prototyping to encompass multiple aspects of station-wide fault management (SWFM) as discussed in OMS requirements documentation.

  9. Automatic braking system modification for the Advanced Transport Operating Systems (ATOPS) Transportation Systems Research Vehicle (TSRV)

    Science.gov (United States)

    Coogan, J. J.

    1986-01-01

    Modifications were designed for the B-737-100 Research Aircraft autobrake system hardware of the Advanced Transport Operating Systems (ATOPS) Program at Langley Research Center. These modifications will allow the on-board flight control computer to control the aircraft deceleration after landing to a continuously variable level for the purpose of executing automatic high speed turn-offs from the runway. A bread board version of the proposed modifications was built and tested in simulated stopping conditions. Test results, for various aircraft weights, turnoff speed, winds, and runway conditions show that the turnoff speeds are achieved generally with errors less than 1 ft/sec.

  10. A continuous winding scheme for superconducting tokamak coils with cable-in-conduit conductor

    International Nuclear Information System (INIS)

    Superconducting magnet coils are essential for steady-state or long-pulse operation of tokamaks. In an advanced tokamak, the central solenoid (CS) coils are usually divided into several pairs of modules to provide for an extra plasma shaping capability in addition to those available from the shaping (poloidal field) coils. In the conventional pancake winding scheme of superconducting coils, each coil consists of separate superconducting 'double-pancake' coils connected together in series; however, such joints are not superconducting, which is one of the major disadvantages, especially in pulsed operations. A new type of winding was adopted for the ITER CS coil, which consists of cylindrical shell 'layers' joined in series. A disadvantage of this layer winding is its inability to yield modular coils that can provide certain degree of plasma shaping. Joints can be removed in a coil winding pack with the conventional pancake winding scheme, if the conductor is sufficiently long and the winding machine is properly equipped. The compactness, however, cannot be preserved with this scheme. The winding compactness is important since the radial build of the CS coils is one of the major parameters that determine the machine size. In this paper, we present a continuous winding scheme that requires no joints, allows coil fabrication at minimum dimension, and meets the flux swing requirement and other practical aspects

  11. Conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR), September 1978

    International Nuclear Information System (INIS)

    The flexibility of the fusion hybrid reactor to function as a fuel production facility, power plant, waste disposal burner or combinations of all of these, as well as the reactor's ability to use proliferation resistant fuel cycles, has provided the incentive to assess the feasibility of a near-term demonstration plant. The goals for a Demonstration Tokamak Hybrid Reactor (DTHR) were established and an initial conceptual design was selected. Reactor performance and economics were evaluated and key developmental issues were assessed. The study has shown that a DTHR is feasible in the late 1980's, a significant quantity of fissile fuel could be produced from fertile thorium using present day fission reactor blanket technology, and a large number of commercially prototypical components and systems could be developed and operationally verified. The DTHR concept would not only serve as proof-of-principle for hybrid technology, but could be operated in the ignited mode and provide major advancements for pure fusion technology

  12. Regolith Advanced Surface Systems Operations Robot (RASSOR) Phase 2 and Smart Autonomous Sand-Swimming Excavator

    Science.gov (United States)

    Sandy, Michael

    2015-01-01

    The Regolith Advanced Surface Systems Operations Robot (RASSOR) Phase 2 is an excavation robot for mining regolith on a planet like Mars. The robot is programmed using the Robotic Operating System (ROS) and it also uses a physical simulation program called Gazebo. This internship focused on various functions of the program in order to make it a more professional and efficient robot. During the internship another project called the Smart Autonomous Sand-Swimming Excavator was worked on. This is a robot that is designed to dig through sand and extract sample material. The intern worked on programming the Sand-Swimming robot, and designing the electrical system to power and control the robot.

  13. Injection of intense ion beam into a tokamak

    International Nuclear Information System (INIS)

    We describe an experiment to investigate the direct injection of an intense ion beam into a tokamak by means of the polarization drift. Confinement of 100 keV ions in the UCI tokamak (r = 15 cm, R = 60 cm, B/sub T/ = 6 kG) requires operation with a plasma current of 56 kA corresponding to q (limiter) = 2. Trapped ions are to be detected by a charge-exchange analyzer. The present status of the experiment will be discussed

  14. Multichannel bolometer for radiation measurements on the TCA tokamak

    International Nuclear Information System (INIS)

    A multichannel radiation bolometer has been developed for the Tokamak Chauffage Alfven (TCA) tokamak. It has 16 equally spaced chords that view the plasma through a narrow horizontal slit. Almost an entire vertical plasma cross section can be observed. The bolometer operates on the basis of a semiconducting element which serves as a temperature-dependent resistance. A new electronic circuit has been developed which takes advantage of the semiconductor characteristics of the detector by using feedback techniques. Measurements made with this instrument are discussed

  15. Spheromak injection into a tokamak

    OpenAIRE

    Brown, M R; Bellan, P. M.

    1990-01-01

    Recent results from the Caltech spheromak injection experiment [to appear in Phys. Rev. Lett.] are reported. First, current drive by spheromak injection into the ENCORE tokamak as a result of the process of magnetic helicity injection is observed. An initial 30% increase in plasma current is observed followed by a drop by a factor of 3 because of sudden plasma cooling. Second, spheromak injection results in an increase of tokamak central density by a factor of 6. The high-current/high-density...

  16. Solenoid-free plasma start-up in spherical tokamaks

    Science.gov (United States)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

  17. Proposals for an influential role of small tokamaks in mainstream fusion physics and technology research

    International Nuclear Information System (INIS)

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confinement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies, which because of the risky nature cannot be done in large machines without preliminary studies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. All these tasks are well recognised and reflected in documents and understood by the large tokamak teams. Recent experimental results will be presented of contributions to mainstream fusion physics and technology research on small tokamaks involved in the IAEA Coordinated Research Project 'Joint Research using small tokamaks', started in 2004

  18. Spherical tokamak research for fusion reactor

    International Nuclear Information System (INIS)

    Between ITER and the commercial fusion reactor, there are many technological problems to be solved such as cost, neutron and steady-state operation. In the conceptual design of VECTOR and Slim CS reactors it was shown that the key is 'low aspect ratio'. The spherical tokamak (ST) has been expected as the base for fusion reactors. In US, ST is considered as a non-superconducting reactor for use in the neutron irradiation facility. Conceptual design of the superconducting ST reactor is conducted in Japan and Korea independently. In the present article, the prospect of the ST reactor design is discussed. (author)

  19. Application of MDSplus on EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    QU Lianzheng; LUO Jiarong; LI lingling; ZHANG Mingxing; WANG Yong

    2007-01-01

    EAST is a fully superconducting Tokamak in China used for controlled fusion research. MDSplus, a special software package for fusion research, has been used successfully as a central repository for analysed data and PCS (Plasma Control System) data since the debugging experiment in the spring of 2006 . In this paper, the reasons for choosing MDSplus as the analysis database and the way to use it are presented in detail, along with the solution to the problem that part of the MDSplus library does not work in the multithread mode. The experiment showed that the data system based on MDSplus operated stably and it could provide a better performance especially for remote users.

  20. Advances in fracture mechanics analyses of primary system performance under operating and accident conditions

    International Nuclear Information System (INIS)

    Safety research sponsored by the Nuclear Regulatory Commission, Division of Reactor Safety Research, has resulted in notable advances in several areas of importance in the safety evaluation of reactor primary systems under normal operations and accident situations. First, the methods of linear elastic fracture mechanics and of elastic plastic fracture mechanics have been validated for prediction of pressure vessel performance by the Intermediate Vessel Test program results at the Oak Ridge National Laboratory. The ability confidently to predict vessel performance under realistic service conditions has permitted development of the computer program OCTAVIA which computes failure curves for a range of flaw sizes in terms of pressure and temperature for specified presure vessel material at specific neutron fluence levels. It then considers the probability of occurrence of flaw sizes and magnitude of pressure during an operational, overpressurization transient and determines the probability of failure, for both individual flaw sizes and for the full spectrum. This advance has been verified by the confirmatory results of testing small thick-walled cylinders under thermal shock conditions in the Heavy Section Steel Technology program, and of warm prestressing tests at the US Navel Research Laboratory. Thirdly, the technology of crack arrest has reached a level wherein standardization of test specimens and testing methods is now possible and, indeed, is underway. (Auth.)

  1. Designing and Operating Through Compromise: Architectural Analysis of CKMS for the Advanced Metering Infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Duren, Mike [Sypris Electronics, LLC; Aldridge, Hal [ORNL; Abercrombie, Robert K [ORNL; Sheldon, Frederick T [ORNL

    2013-01-01

    Compromises attributable to the Advanced Persistent Threat (APT) highlight the necessity for constant vigilance. The APT provides a new perspective on security metrics (e.g., statistics based cyber security) and quantitative risk assessments. We consider design principals and models/tools that provide high assurance for energy delivery systems (EDS) operations regardless of the state of compromise. Cryptographic keys must be securely exchanged, then held and protected on either end of a communications link. This is challenging for a utility with numerous substations that must secure the intelligent electronic devices (IEDs) that may comprise complex control system of systems. For example, distribution and management of keys among the millions of intelligent meters within the Advanced Metering Infrastructure (AMI) is being implemented as part of the National Smart Grid initiative. Without a means for a secure cryptographic key management system (CKMS) no cryptographic solution can be widely deployed to protect the EDS infrastructure from cyber-attack. We consider 1) how security modeling is applied to key management and cyber security concerns on a continuous basis from design through operation, 2) how trusted models and key management architectures greatly impact failure scenarios, and 3) how hardware-enabled trust is a critical element to detecting, surviving, and recovering from attack.

  2. Commissioning of the University of Maryland Electron Ring (UMER): Advances toward multiturn operation

    International Nuclear Information System (INIS)

    The University of Maryland Electron Ring (UMER) is a low-energy, high current recirculator for beam physics research [M. Reiser et al., in Proceedings of the 1999 Particle Accelerator Conference, New York, NY (IEEE, New York, 1999), p. 234]. Ring construction is completed for multiturn operation of beams over a broad range of intensities and initial conditions. UMER is an extremely versatile experimental platform with a beam current of up to 100 mA and a pulse length as long as 100 ns. UMER is addressing issues in beam physics relevant to many applications that require intense beams of high quality, such as advanced concept accelerators, free electron lasers, spallation neutron sources, and future heavy-ion drivers for inertial fusion. The primary focus of this paper is to present experimental results in the areas of beam steering and multiturn operation of the ring. Unique beam steering algorithms now include measurement of the beam response matrix at each quadrupole and matrix inversion by singular value decomposition. With these advanced steering methods, transport of an intense beam over four turns (144 full lattice periods) of the ring has been achieved

  3. Advanced satellite workstation: An integrated workstation environment for operational support of satellite system planning and analysis

    Science.gov (United States)

    Sutton, Stewart A.

    1992-01-01

    A prototype integrated environment, the Advanced Satellite Workstation (ASW), is described that has been developed and delivered for evaluation and operator feedback in an operational satellite control center. The current ASW hardware consists of a Sun Workstation and Macintosh II Workstation connected via an ethernet Network Hardware and Software, Laser Disk System, Optical Storage System, and Telemetry Data File Interface. The central mission of ASW is to provide an intelligent decision support and training environment for operator/analysts of complex systems such as satellites. There have been many workstation implementations recently which incorporate graphical telemetry displays and expert systems. ASW is a considerably broader look at intelligent, integrated environments for decision support, based upon the premise that the central features of such an environment are intelligent data access and integrated toolsets. A variety of tools have been constructed in support of this prototype environment including: an automated pass planner for scheduling vehicle support activities, architectural modeler for hierarchical simulation and analysis of satellite vehicle subsystems, multimedia-based information systems that provide an intuitive and easily accessible interface to Orbit Operations Handbook and other relevant support documentation, and a data analysis architecture that integrates user modifiable telemetry display systems, expert systems for background data analysis, and interfaces to the multimedia system via inter-process communication.

  4. A comparison of laparoscopic and open surgery following pre-operative chemoradiation therapy for locally advanced lower rectal cancer

    International Nuclear Information System (INIS)

    Although pre-operative chemoradiation therapy for advanced lower rectal cancer is a controversial treatment modality, it is increasingly used in combination with surgery. Few studies have considered the combination of chemoradiation therapy followed by laparoscopic surgery for locally advanced lower rectal cancer; therefore, this study aimed to assess the usefulness of this therapeutic combination. We retrospectively reviewed the medical records of patients with locally advanced lower rectal cancer treated by pre-operative chemoradiation therapy and surgery from February 2002 to November 2012 at Oita University. We divided patients into an open surgery group and a laparoscopic surgery group and evaluated various parameters by univariate and multivariate analyses. In total, 33 patients were enrolled (open surgery group, n=14; laparoscopic surgery group, n=19). Univariate analysis revealed that compared with the open surgery group, operative time was significantly longer, whereas intra-operative blood loss and intra-operative blood transfusion requirements were significantly less in the laparoscopic surgery group. There were no significant differences in post-operative complication and recurrence rates between the two groups. According to multivariate analysis, operative time and intra-operative blood loss were significant predictors of outcome in the laparoscopic surgery group. This study suggests that laparoscopic surgery after chemoradiation therapy for locally advanced lower rectal cancer is a safe procedure. Further prospective investigation of the long-term oncological outcomes of laparoscopic surgery after chemoradiation therapy for locally advanced lower rectal cancer is required to confirm the advantages of laparoscopic surgery over open surgery. (author)

  5. Alternate Data Acquisition and Real-time Monitoring System on HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    Wei Peijie; Luo Jiarong; Wang Hua; Li Guiming

    2005-01-01

    A new system called alternate data acquisition and real-time monitoring system has been developed for long-time discharge in tokamak operation. It can support continuous on-line data acquisition at a high sampling rate and a graphic display of the plasma parameters during the discharge. Thus operators can monitor and control the plasma state in real time. An application of this system has been demonstrated on the HT-7 tokamak.

  6. Full Tokamak discharge simulation and kinetic plasma profile control for ITER

    International Nuclear Information System (INIS)

    Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

  7. Final Report - Advanced MEA's for Enhanced Operating Conditions, Amenable to High Volume Manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Debe, Mark K.

    2007-09-30

    This report summarizes the work completed under a 3M/DOE contract directed at advancing the key fuel cell (FC) components most critical for overcoming the polymer electrolyte membrane fuel cell (PEMFC) performance, durability & cost barriers. This contract focused on the development of advanced ion exchange membranes & electrocatalysts for PEMFCs that will enable operation under ever more demanding automotive operating conditions & the use high volume compatible processes for their manufacture. Higher performing & more durable electrocatalysts must be developed for PEMFCs to meet the power density & lifetime hours required for FC vehicles. At the same time the amount of expensive Pt catalyst must be reduced to lower the MEA costs. While these two properties are met, the catalyst must be made resistant to multiple degradation mechanisms to reach necessary operating lifetimes. In this report, we present the work focused on the development of a completely new approach to PEMFC electrocatalyts, called nanostructured thin film (NSTF) catalysts. The carbon black supports are eliminated with this new approach which eliminates the carbon corrosion issue. The thin film nature of the catalyst significantly improves its robustness against dissolution & grain growth, preserving the surface area. Also, the activity of the NSTF for oxygen reduction is improved by over 500% compared to dispersed Pt catalyts. Finally, the process for fabricating the NSTF catalysts is consistent with high volume roll-good manufacturing & extremely flexible towards the introduction of new catalyst compositions & structures. This report documents the work done to develop new multi-element NSTF catalysts with properties that exceed pure Pt, that are optimized for use with the membranes discussed below, & advance the state-of-the-art towards meeting the DOE 2010 targets for PEMFC electrocatalysts. The work completed advances the understanding of the NSTF catalyst technology, identifies new NSTF

  8. Conceptual tokamak design at high neutron fluence

    International Nuclear Information System (INIS)

    For the future fusion reactor, it is important to design an experimental device that can be performed testing in-vessel components including tritium breeding modules relevant to the future fusion reactor with high neutron fluence. To realize this requirement, a conceptual tokamak design has been performed in accordance with plasma performance and shape at quasi-steady-state operation. One of the promising scenarios for this purpose is proposed to produce the plasma at the outward shifted radial position with a small minor radius for reasonable plasma parameters. From the analytical results, an appropriate space can be found for neutron shielding so that additional neutron shielding can be installed to protect the tokamak components from any neutron damages under the neutron fluence of 1 MWa m-2. Based on the structural analyses, a two-stage blanket module concept is proposed, i.e. one shielding block with the first wall assembly during high Q operation and two shielding blocks or additional tritium breeding modules during quasi-steady state operation

  9. Tokamak Scenario Trajectory Optimization Using Fast Integrated Simulations

    Science.gov (United States)

    Urban, Jakub; Artaud, Jean-François; Vahala, Linda; Vahala, George

    2015-11-01

    We employ a fast integrated tokamak simulator, METIS, for optimizing tokamak discharge trajectories. METIS is based on scaling laws and simplified transport equations, validated on existing experiments and capable of simulating a full tokamak discharge in about 1 minute. Rapid free-boundary equilibrium post-processing using FREEBIE provides estimates of PF coil currents or forces. We employ several optimization strategies for optimizing key trajectories, such as Ip or heating power, of a model ITER hybrid discharge. Local and global algorithms with single or multiple objective functions show how to reach optimum performance, stationarity or minimum flux consumption. We constrain fundamental operation parameters, such as ramp-up rate, PF coils currents and forces or heating power. As an example, we demonstrate the benefit of current over-shoot for hybrid mode, consistent with previous results. This particular optimization took less than 2 hours on a single PC. Overall, we have established a powerful approach for rapid, non-linear tokamak scenario optimization, including operational constraints, pertinent to existing and future devices design and operation.

  10. Improvement of environmental aspects of thermal power plant operation by advanced control concepts

    Directory of Open Access Journals (Sweden)

    Mikulandrić Robert

    2012-01-01

    Full Text Available The necessity of the reduction of greenhouse gas emissions, as formulated in the Kyoto Protocol, imposes the need for improving environmental aspects of existing thermal power plants operation. Improvements can be reached either by efficiency increment or by implementation of emission reduction measures. Investments in refurbishment of existing plant components or in plant upgrading by flue gas desulphurization, by primary and secondary measures of nitrogen oxides reduction, or by biomass co-firing, are usually accompanied by modernisation of thermal power plant instrumentation and control system including sensors, equipment diagnostics and advanced controls. Impact of advanced control solutions implementation depends on technical characteristics and status of existing instrumentation and control systems as well as on design characteristics and actual conditions of installed plant components. Evaluation of adequacy of implementation of advanced control concepts is especially important in Western Balkan region where thermal power plants portfolio is rather diversified in terms of size, type and commissioning year and where generally poor maintenance and lack of investments in power generation sector resulted in high greenhouse gases emissions and low efficiency of plants in operation. This paper is intended to present possibilities of implementation of advanced control concepts, and particularly those based on artificial intelligence, in selected thermal power plants in order to increase plant efficiency and to lower pollutants emissions and to comply with environmental quality standards prescribed in large combustion plant directive. [Acknowledgements. This paper has been created within WBalkICT - Supporting Common RTD actions in WBCs for developing Low Cost and Low Risk ICT based solutions for TPPs Energy Efficiency increasing, SEE-ERA.NET plus project in cooperation among partners from IPA SA - Romania, University of Zagreb - Croatia and Vinca

  11. Real-time control of current and pressure profiles in tokamak plasmas

    International Nuclear Information System (INIS)

    Recent progress in the field of 'advanced tokamak scenarios' prefigure the operation regime of a future thermonuclear fusion power plant. Compared to the reference regime, these scenarios offer a longer plasma confinement time thanks to increased magnetohydrodynamic stability and to a better particle and energy confinement through a reduction of plasma turbulence. This should give access to comparable fusion performances at reduced plasma current and could lead to a steady state fusion reactor since the plasma current could be entirely generated non-inductively. Access to this kind of regime is provided by the existence of an internal transport barrier, linked to the current profile evolution in the plasma, which leads to steep temperature and pressure profiles. The comparison between heat transport simulations and experiments allowed the nature of the barriers to be better understood as a region of strongly reduced turbulence. Thus, the control of this barrier in a stationary manner would be a remarkable progress, in particular in view of the experimental reactor ITER. The Tore Supra and JET tokamaks, based in France and in the United Kingdom, constitute ideal instruments for such experiments: the first one allows stationary plasmas to be maintained during several minutes whereas the second one provides unique fusion performances. In Tore Supra, real-time control experiments have been accomplished where the current profile width and the pressure profile gradient were controlled in a stationary manner using heating and current drive systems as actuators. In the JET tokamak, the determination of an empirical static model of the plasma allowed the current and pressure profiles to be simultaneously controlled and so an internal transport barrier to be sustained. Finally, the identification of a dynamic model of the plasma led to the definition of a new controller capable, in principle, of a more efficient control. (author)

  12. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  13. Expanding Robust HCCI Operation with Advanced Valve and Fuel Control Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Szybist, J. P. [Oak Ridge National Lab., Oak Ridge, TN (United States); Confer, K. [Delphi Automotive Systems (United States)

    2012-09-11

    Delphi Automotive Systems and ORNL established this CRADA to advance the commercialization potential of the homogeneous charge compression ignition (HCCI) advanced combustion strategy for gasoline engine platforms. HCCI combustion has been shown by others to produce high diesel-like efficiency on a gasoline engine platform while simultaneously producing low NOX and particulate matter emissions. However, the commercialization barriers that face HCCI combustion are significant, with requirements for a more active engine control system, likely with next-cycle closed-loop feedback control, and with advanced valve train technologies to enable negative valve overlap conditions. In the partnership between Delphi and ORNL, each organization brought a unique and complementary set of skills to the project. Delphi has made a number of breakthroughs with production-intent valve train technologies and controls in recent years to make a part time production-intent HCCI engine plausible. ORNL has extensive knowledge and expertise with HCCI combustion, and also has a versatile research engine with hydraulic valve actuation (HVA) that is useful for guiding production of a cam-based HCCI system. Partnering these knowledge bases and capabilities was essential towards making progress to better understand HCCI combustion and the commercialization barriers that it faces. ORNL and Delphi maintained strong collaboration throughout the project. Meetings were held regularly, with additional reports, presentations, and meetings as necessary to maintain progress. Delphi provided guidance to ORNL regarding operational strategies to investigate on their single-cylinder research engine with HVA and data from their experimental multi-cylinder engine for modeling. ORNL provided single-cylinder engine data and modeling results.

  14. Operational experiences in MOX fuel fabrication for the FUGEN advanced thermal reactor

    International Nuclear Information System (INIS)

    The Japan Nuclear Cycle Development Institute, JNC, has fabrication the MOX fuel for the Advanced Thermal Reactor, ATR, ''FUGEN'' in the Plutonium Fuel Fabrication Facility, PFFF, since 1974. For these 25 years, the MOX fuel fabrication has progressed in stable manner after overcoming several problems at the start up of FUGEN fuel fabrication. Through the experience, improvements on process equipment and conditions have been taken place to achieve efficient MOX fuel fabrication on an engineering scale as 10 tons MOX per year. Main features of current fabrication process are digested as one step blending with ball milling, pelletizing without granulation and sintering with batch type furnaces. This fabrication process has been demonstrated and confirmed to be applicable techniques for the MOX fuel fabrication on this scale. This paper discusses the FUGEN fuel fabrication focused on the MOX pellet fabrication with operational experiences and improvements to the process. (author)

  15. OPERATION OF ENTERPRISE RESOURCE PLANNING SYSTEM IMPLEMENTATION COMPATIBILITY TOWARDS TECHNICAL ADVANCEMENT

    Directory of Open Access Journals (Sweden)

    J. Venkatesh

    2013-01-01

    Full Text Available The Execution of ERP systems has been perplexing factor for many firms. Many establishments have accepted that the operation of ERP system is a massive hindrance until the flow of process is organized carefully. As information technology is booming its prerequisite for organizations to realize the prominence of technical advancement and compatibility in work environment. A complete review was done to ascertain the features and strategic aids of ERP enactments using the retorts from 120 firms. The respondents were approached with orderly framed questionnaires, thereby giving them ample time to come out with their own thoughts. The effects of this learning provide assistance for the vendors, higher officials and ERP specialists to be more competent in handling the execution of ERP with their inadequate possessions there by augmenting the business. It acts a pathway for the concerns to realize their potent and extend their business platform.

  16. Advanced Transport Operating System (ATOPS) color displays software description microprocessor system

    Science.gov (United States)

    Slominski, Christopher J.; Plyler, Valerie E.; Dickson, Richard W.

    1992-01-01

    This document describes the software created for the Sperry Microprocessor Color Display System used for the Advanced Transport Operating Systems (ATOPS) project on the Transport Systems Research Vehicle (TSRV). The software delivery known as the 'baseline display system', is the one described in this document. Throughout this publication, module descriptions are presented in a standardized format which contains module purpose, calling sequence, detailed description, and global references. The global reference section includes procedures and common variables referenced by a particular module. The system described supports the Research Flight Deck (RFD) of the TSRV. The RFD contains eight cathode ray tubes (CRTs) which depict a Primary Flight Display, Navigation Display, System Warning Display, Takeoff Performance Monitoring System Display, and Engine Display.

  17. Advanced moisture separation and reheat systems: 20 years of operating experience

    International Nuclear Information System (INIS)

    Following the world oil crisis of the mid-seventies, the French utility Electricite de France (EDF) commissioned its first 900 MWe PWR at Fessenheim 1, in December 1977, i.e. more than twenty years ago. For such nuclear power stations, the steam Moisture Separation and Reheat systems, which ensure high pressure turbine shell exhaust steam drying and superheat prior to expansion into the low (or intermediate) pressure shells, are key components as far as steam plant performance and reliability are concerned. The paper summarizes the operating experience gained for such systems designed and supplied by GEC ALSTHOM on four continents. The paper presents as well the advanced double stage reheat Moisture Separator Reheater concept, introduced by GEC ALSTHOM Stein Industrie ten years ago. This design has been successfully implemented for the French 1500 MWe GEC ALSTHOM ArabelleTM steam turbines, such as those equipping the award-winning Chooz B nuclear power station, commissioned in 1997. (author)

  18. Collagen-based biological glue after Appleby operation for advanced gastric cancer

    Institute of Scientific and Technical Information of China (English)

    Gianluca Baiocchi; Nazario Portolani; Federico Gheza; Stefano M Giulini

    2011-01-01

    Pancreatic fistula is a common complication of distal pancreatectomy; although various surgical procedures have been proposed, no clear advantage is evident for a single technique. We herein report the case of a 38-year-old patient affected by an advanced gastric carcinoma infiltrating the pancreas body, with extensive nodal metastases involving the celiac trunk, who underwent total gastrectomy with lymphadenectomy, distal pancreatectomy and resection en bloc of the celiac trunk (Appleby operation). At the end of the demolitive phase, thepancreaticstumpandtheaortahe pancreatic stump and the aorta at the level of the celiac ligature were covered with a layer of Tachosil(R), ahorsecollagenspongemadewitha horse collagen sponge made with human coagulation factors (fibrinogen and thrombin). Presenting this case, we wish to highlight the possible sealing effect of this product and hypothesize a role in preventing pancreatic fistula and postoperative lymphorrhagia from extensive nodal dissection.

  19. Advanced Transport Operating System (ATOPS) color displays software description: MicroVAX system

    Science.gov (United States)

    Slominski, Christopher J.; Plyler, Valerie E.; Dickson, Richard W.

    1992-01-01

    This document describes the software created for the Display MicroVAX computer used for the Advanced Transport Operating Systems (ATOPS) project on the Transport Systems Research Vehicle (TSRV). The software delivery of February 27, 1991, known as the 'baseline display system', is the one described in this document. Throughout this publication, module descriptions are presented in a standardized format which contains module purpose, calling sequence, detailed description, and global references. The global references section includes subroutines, functions, and common variables referenced by a particular module. The system described supports the Research Flight Deck (RFD) of the TSRV. The RFD contains eight Cathode Ray Tubes (CRTs) which depict a Primary Flight Display, Navigation Display, System Warning Display, Takeoff Performance Monitoring System Display, and Engine Display.

  20. Advanced Transport Operating System (ATOPS) Flight Management/Flight Controls (FM/FC) software description

    Science.gov (United States)

    Wolverton, David A.; Dickson, Richard W.; Clinedinst, Winston C.; Slominski, Christopher J.

    1993-01-01

    The flight software developed for the Flight Management/Flight Controls (FM/FC) MicroVAX computer used on the Transport Systems Research Vehicle for Advanced Transport Operating Systems (ATOPS) research is described. The FM/FC software computes navigation position estimates, guidance commands, and those commands issued to the control surfaces to direct the aircraft in flight. Various modes of flight are provided for, ranging from computer assisted manual modes to fully automatic modes including automatic landing. A high-level system overview as well as a description of each software module comprising the system is provided. Digital systems diagrams are included for each major flight control component and selected flight management functions.

  1. The Chimera II Real-Time Operating System for advanced sensor-based control applications

    Science.gov (United States)

    Stewart, David B.; Schmitz, Donald E.; Khosla, Pradeep K.

    1992-01-01

    Attention is given to the Chimera II Real-Time Operating System, which has been developed for advanced sensor-based control applications. The Chimera II provides a high-performance real-time kernel and a variety of IPC features. The hardware platform required to run Chimera II consists of commercially available hardware, and allows custom hardware to be easily integrated. The design allows it to be used with almost any type of VMEbus-based processors and devices. It allows radially differing hardware to be programmed using a common system, thus providing a first and necessary step towards the standardization of reconfigurable systems that results in a reduction of development time and cost.

  2. Complete rank theorem of advanced calculus and singularities of bounded linear operators

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Let E and F be Banach spaces, f: U E→ F be a map of Cr (r≥ 1), x∈ U, and f'(x0) denote the Fréchet differential of f at x0. Suppose that f'(xo) is double split, Rank(f'(x0)) = ∞, dimN(f'(x0)) > 0 and codimR(f'(x0)) > 0. The rank theorem in advanced calculus asks to answer what properties of f ensure that f(x) is conjugate to f'(x0) near x0. We have proved that the conclusion of the theorem is equivalent to one kind of singularities for bounded linear operators, I.e., x0 is a locally fine point for f'(x) or generalized regular point of f(x); so, a complete rank theorem in advanced calculus is established, I.e., a sufficient and necessary condition such that the conclusion of the theorem to be held is given.

  3. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  4. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  5. Development of a test bed for operator aid and advanced control concepts in nuclear power plants

    International Nuclear Information System (INIS)

    A great amount of research and development is currently under way in the utilization of artificial intelligence (AI), expert system, and control theory advances in nuclear power plants as a basis for operator aids and automatic control systems. This activity requires access to the measured dynamic responses of the plant to malfunction, operator- or automatic-control-initiated actions. This can be achieved by either simulating plant behavior or by using an actual plant. The advantage of utilizing an actual plant versus a simulator is that the true behavior is assured of both the power generation system and instrumentation. Clearly, the disadvantages of using an actual plant are availability due to licensing, economic, and risk constraints and inability to address accident conditions. In this work the authors have decided to employ a functional one-ninth scale model of a pressurized water reactor (PWR). The scaled PWR (SPWR) facility is a two-loop representation of a Westinghouse PWR utilizing freon as the working fluid and electric heater rods for the core. The heater rods are driven by a neutron kinetics model accounting for measured thermal core conditions. A control valve in the main steam line takes the place of the turbine generator. A range of normal operating and accident situations can be addressed. The SPWR comes close to offering all the advantages of both a simulator and an actual physical plant in regard to research and development on AI, expert system, and control theory applications. The SPWR is being employed in the development of an expert-system-based operator aid system. The current status of this project is described

  6. A Multifaceted Approach to Modernizing NASA's Advanced Multi-Mission Operations System (AMMOS) System Architecture

    Science.gov (United States)

    Estefan, Jeff A.; Giovannoni, Brian J.

    2014-01-01

    The Advanced Multi-Mission Operations Systems (AMMOS) is NASA's premier space mission operations product line offering for use in deep-space robotic and astrophysics missions. The general approach to AMMOS modernization over the course of its 29-year history exemplifies a continual, evolutionary approach with periods of sponsor investment peaks and valleys in between. Today, the Multimission Ground Systems and Services (MGSS) office-the program office that manages the AMMOS for NASA-actively pursues modernization initiatives and continues to evolve the AMMOS by incorporating enhanced capabilities and newer technologies into its end-user tool and service offerings. Despite the myriad of modernization investments that have been made over the evolutionary course of the AMMOS, pain points remain. These pain points, based on interviews with numerous flight project mission operations personnel, can be classified principally into two major categories: 1) information-related issues, and 2) process-related issues. By information-related issues, we mean pain points associated with the management and flow of MOS data across the various system interfaces. By process-related issues, we mean pain points associated with the MOS activities performed by mission operators (i.e., humans) and supporting software infrastructure used in support of those activities. In this paper, three foundational concepts-Timeline, Closed Loop Control, and Separation of Concerns-collectively form the basis for expressing a set of core architectural tenets that provides a multifaceted approach to AMMOS system architecture modernization intended to address the information- and process-related issues. Each of these architectural tenets will be further explored in this paper. Ultimately, we envision the application of these core tenets resulting in a unified vision of a future-state architecture for the AMMOS-one that is intended to result in a highly adaptable, highly efficient, and highly cost

  7. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  8. Compact ignition tokamak physics and engineering basis

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is a high-field, compact tokamak design whose objective is the study of physics issues associated with burning plasmas. The toroidal and poloidal field coils employ a copper-steel laminate, manufactured by explosive-bonding techniques, to support the forces generated by the design fields: 10 T toroidal field at the plasma center; 21 T in the OH solenoid. A combination of internal and external PF coils provides control of the equilibrium and the ability to sweep the magnetic separatrix across the divertor plates during a pulse. At temperatures and βα levels characteristic of ITER designs, the fusion power in CIT approaches 800 MW and can be the limiting factor in the pulse length. Ignition requires that the confinement time exceed present L-mode scalings by about a factor of two, which is anticipated to occur as a result of the operational flexibility incorporated into the design. Conventional operating limits given by 20 e and qψ ≤ 3.2 have been chosen and, in the case of MHD limits, have been justified by ideal stability analysis. The power required for CIT ignition ranges from 10 MW to 40 MW or more, depending on confinement assumptions, and either ICRF or ECRF heating, or both, will be used. (author). 17 refs, 6 figs, 1 tab

  9. 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.

  10. 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.

  11. Performance of a New Ion Source for KSTAR Tokamak Plasma Heating

    Science.gov (United States)

    Tae-Seong, Kim; Seung, Ho Jeong; Doo, Hee Chang; Kwang, Won Lee; Sang-Ryul, In

    2014-06-01

    In the experimental campaign of 2010 and 2011 on KSTAR, the NBI-1 system was equipped with one prototype ion source and operated successfully, providing a neutral beam power of 0.7-1.6 MW to the tokamak plasma. The new ion source planned for the 2012 KSTAR campaign had a much more advanced performance compared with the previous one. The target performance of the new ion source was to provide a neutral deuterium beam of 2 MW to the tokamak plasma. The ion source was newly designed, fabricated, and assembled in 2011. The new ion source was then conditioned up to 64 A/100 keV over a 2-hour beam extraction and performance tested at the NB test stand (NBTS) at the Korea Atomic Energy Research Institute (KAERI) in 2012. The measured optimum perveance at which the beam divergence is a minimum was about 2.5 μP, and the minimum beam divergent angle was under 1.0° at 60 keV. These results indicate that the 2.0 MW neutral beam power at 100 keV required for the heating of plasma in KSTAR can be delivered by the installation of the new ion source in the KSTAR NBI-1 system.

  12. Lithium capillary porous system behavior as PFM in FTU Tokamak experiments

    International Nuclear Information System (INIS)

    Full text of publication follows: Liquid lithium use on the base of capillary porous systems (CPS) application as plasma facing material (PFM) of tokamaks is advanced way to solve the problems of plasma contamination with high Z impurity, PFM degradation and tritium retention. In frame of joint program between ENEA (Italy) and FSUE 'Red Star' and TRINITI (RF) started at the end of 2005 die test of liquid lithium limiter (LLL) with CPS in a high field, medium size, carbon free tokamak FTU have been performed successfully. The LLL has been inserted in ohmic plasma discharges and at additional heating with LH and ECR at power levels in the MW range without any particular problem (BT = 6 T, Ip = 0.5- 0.9 MA, ne = 0.2 -2.6x1020 m-3, t = 1.5 s, P∼ 2-5 MW/m2 at a normal discharge). The behavior of lithium CPS based on stainless steel wire mesh and its surface modification in normal discharges and at disruptions has been studied. Results of microscopic analyses of CPS structure after experimental campaigns are presented. The possibility to withstand heat load exceeding 5 MW/m2 without damage, lithium surface renewal, mechanical stabilization of liquid lithium against MHD forces have been confirmed. Application of W, Mo as the base material and possible structure types of CPS have been considered for operating parameters improvement of long-living plasma facing components. (authors)

  13. Collection and Characterization of Particulate from the Tore Supra Tokamak (Dec. 1999 Vent)

    Energy Technology Data Exchange (ETDEWEB)

    Sharpe, John Phillip

    2002-12-01

    Particulate generated during the operation of a fusion device contributes to the radiological source term associated with accident scenarios in the device.1,2 Understanding the mechanisms generating the particulate and correctly describing its physical and chemical behavior is essential for safety analyses of future fusion devices. Knowledge of these mechanisms is gained by collecting and characterizing particulate matter from operating fusion facilities. Tokamak dust, the particulate matter generated during the operation of a tokamak fusion device, was collected from Tore Supra in December 1999, during the initial phase of the scheduled shutdown for installation of advanced plasma facing components. Tore Supra, located at CEA Cadarache, France, is presently the third largest operating tokamak with the capability of long-pulse operation. Eighteen super-conducting coils produce the 4.5 T magnetic field inside a vessel 2.4 m in major radius and 1.2 m in minor radius. Limiter and divertor regimes of operation are possible. In the divertor regime, the circular magnetic configuration is ergodized by six outboard resonant divertor modules that are covered with 12 m2 of carbon fiber composite (CFC) tiles. In addition, some field lines are diverted to actively cooled neutralizing plates made of CuCrZr bars covered with B4C. In the limiter regime, the plasma leans on the actively cooled inboard first wall or on a set of inertially cooled pumped limiters. The first wall area of 12 m2 is covered with both polycrystalline graphite tiles (83%) and CFC tiles (17%). The single outboard limiter is constructed of pyrolitic graphite, and the four toroidally symmetric bottom limiters are constructed of CFC. Figure 1.1 displays the relative location of plasma facing components within the plasma chamber of Tore Supra. In this report, we present in Section 2 the methods used to collect and analyze this dust and describe the selection of sampling locations. Section 3 includes a

  14. Experimental progress and innovation on the HL-2A tokamak

    International Nuclear Information System (INIS)

    The HL-2A Tokamak is the first large controlled fusion experiment device with divertor in China. In this paper, the main experimental results on this device will be presented. Since its establishment, the operation conditions have been improved greatly. First, the divertor configuration was realized, the electron temperature was increased to 5.5 keV by electron cyclotron heating, and then the high confinement edge localized mode was achieved. With the development of high power auxiliary heating and advanced plasma diagnostic systems,innovative contributions have been made in certain plasma physics areas. The 3-dimension structure of the zonal flow has been identified, which is very important in the transport of fusion plasma. Supersonic molecular beam injection has also been developed and successfully used for plasma transport. The tearing mode has been suppressed by electron cyclotron resonance heating with low frequency modulation, and the confinement has been improved. The new phenomena of the internal kink mode and Alfen mode excited by energetic electrons have been observed. Future plans and new experiments on the device will also be briefly presented. (authors)

  15. Future directions in fusion research: Super high field tokamaks

    International Nuclear Information System (INIS)

    Recent experimental results and advances in magnet engineering suggest that super high field, high aspect ratio tokamak devices could be a very efficient way to achieve burning plasma conditions and could open up a new area of research. Copper magnet devices with fields of 13 to 25 T at the plasma are considered. The super high field approach could also provide advantages for ETR and demonstration/commercial reactor concepts (magnetic fields at the plasma in the 8 to 13 T range)

  16. Development of Alfven wave antenna system for TCABR Tokamak

    International Nuclear Information System (INIS)

    The advanced antenna system for Alfven wave plasma heating and current drive in TCABR tokamak is presented. The antenna system is capable of exciting the travelling waves M=- 1, N=-4, -6 with single helicity and provides the possibility to improve Alfven wave plasma heating efficiency and to increase RF power input up 1 MW, without an uncontrolled density rise. The basic features of the antenna design and the results of preliminary tests are analyzed. (author)

  17. Assessment of geometric errors of Advanced Himawari-8 Imager (AHI) over one year operation

    Science.gov (United States)

    Takeuchi, Wataru

    2016-06-01

    This paper presents an approach to check a geometric performance of Advanced Himawari-8 imager (AHI) and demonstrate and evaluate a new approach to ensure more geometric accurately focusing on visible imagery in 500 meters. A series of processing is supplemented by ground control points of shore lines, land mark locations and digital elevation model. Firstly, a template matching technique is conducted to find a best matching point by simply moving the center of AHI sub-image over each point in a reference image of shore lines and calculating the sum of products between the coefficients and the corresponding neighbourhood pixels in the area spanned by the filter mask. Secondly, ortho-rectification processing is carried out to compensate for the geodetical distortions with respect to the acquisition condition including viewing geometry and so on. As a result, an average of root mean square sum of residual errors with system correction and that of precise geometric correction are shown. Overall geometric accuracy is about 1 to 1.5 pixels from 2015 March to July and it also gradually decreased down to 0.2 to 0.8 from 2015 September to 2016 February. AHI is officially open to public for operational use as of July 1, 2015 and after that operation date geometric errors are reasonably satisfied within one pixels of errors.

  18. NASA's Advanced Multimission Operations System: A Case Study in Formalizing Software Architecture Evolution

    Science.gov (United States)

    Barnes, Jeffrey M.

    2011-01-01

    All software systems of significant size and longevity eventually undergo changes to their basic architectural structure. Such changes may be prompted by evolving requirements, changing technology, or other reasons. Whatever the cause, software architecture evolution is commonplace in real world software projects. Recently, software architecture researchers have begun to study this phenomenon in depth. However, this work has suffered from problems of validation; research in this area has tended to make heavy use of toy examples and hypothetical scenarios and has not been well supported by real world examples. To help address this problem, I describe an ongoing effort at the Jet Propulsion Laboratory to re-architect the Advanced Multimission Operations System (AMMOS), which is used to operate NASA's deep-space and astrophysics missions. Based on examination of project documents and interviews with project personnel, I describe the goals and approach of this evolution effort and then present models that capture some of the key architectural changes. Finally, I demonstrate how approaches and formal methods from my previous research in architecture evolution may be applied to this evolution, while using languages and tools already in place at the Jet Propulsion Laboratory.

  19. The CIT [compact ignition tokamak] pellet injection system: Description and supporting research and development

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) will use an advance, high-velocity pellet injection system to achieve and maintain ignited plasmas. Two pellet injectors are provided: a moderate-velocity (1-to 1.5-km/s), single-stage pneumatic injector with high reliability and a high-velocity (4- to 5-km/s), two-stage pellet injector that uses frozen hydrogenic pellets encased in sabots. Both pellet injectors are qualified for operation with tritium feed gas. Issues such as performance, neutron activation of injector components, maintenance, design of the pellet injection vacuum line, gas loads to the reprocessing system, and equipment layout are discussed. Results and plans for supporting research and development (R and D) in the areas of tritium pellet fabrication and high-velocity, repetitive two-stage pneumatic injectors are presented. 7 refs., 4 figs., 2 tabs

  20. Final Report - Advanced MEA's for Enhanced Operating Conditions, Amenable to High Volume Manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Debe, Mark K.

    2007-09-30

    This report summarizes the work completed under a 3M/DOE contract directed at advancing the key fuel cell (FC) components most critical for overcoming the polymer electrolyte membrane fuel cell (PEMFC) performance, durability & cost barriers. This contract focused on the development of advanced ion exchange membranes & electrocatalysts for PEMFCs that will enable operation under ever more demanding automotive operating conditions & the use high volume compatible processes for their manufacture. Higher performing & more durable electrocatalysts must be developed for PEMFCs to meet the power density & lifetime hours required for FC vehicles. At the same time the amount of expensive Pt catalyst must be reduced to lower the MEA costs. While these two properties are met, the catalyst must be made resistant to multiple degradation mechanisms to reach necessary operating lifetimes. In this report, we present the work focused on the development of a completely new approach to PEMFC electrocatalyts, called nanostructured thin film (NSTF) catalysts. The carbon black supports are eliminated with this new approach which eliminates the carbon corrosion issue. The thin film nature of the catalyst significantly improves its robustness against dissolution & grain growth, preserving the surface area. Also, the activity of the NSTF for oxygen reduction is improved by over 500% compared to dispersed Pt catalyts. Finally, the process for fabricating the NSTF catalysts is consistent with high volume roll-good manufacturing & extremely flexible towards the introduction of new catalyst compositions & structures. This report documents the work done to develop new multi-element NSTF catalysts with properties that exceed pure Pt, that are optimized for use with the membranes discussed below, & advance the state-of-the-art towards meeting the DOE 2010 targets for PEMFC electrocatalysts. The work completed advances the understanding of the NSTF catalyst technology, identifies new NSTF

  1. An empirical study on the basic human error probabilities for NPP advanced main control room operation using soft control

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Inseok, E-mail: nuclear82@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Kim, Ar Ryum, E-mail: arryum@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Harbi, Mohamed Ali Salem Al, E-mail: 100035556@kustar.ac.ae [Department of Nuclear Engineering, Khalifa University of Science, Technology and Research, P.O. Box 127788, Abu Dhabi (United Arab Emirates); Lee, Seung Jun, E-mail: sjlee@kaeri.re.kr [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, 150-1, Dukjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Seong, Poong Hyun, E-mail: phseong@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2013-04-15

    Highlights: ► The operation environment of MCRs in NPPs has changed by adopting new HSIs. ► The operation action in NPP Advanced MCRs is performed by soft control. ► Different basic human error probabilities (BHEPs) should be considered. ► BHEPs in a soft control operation environment are investigated empirically. ► This work will be helpful to verify if soft control has positive or negative effects. -- Abstract: By adopting new human–system interfaces that are based on computer-based technologies, the operation environment of main control rooms (MCRs) in nuclear power plants (NPPs) has changed. The MCRs that include these digital and computer technologies, such as large display panels, computerized procedures, soft controls, and so on, are called Advanced MCRs. Among the many features in Advanced MCRs, soft controls are an important feature because the operation action in NPP Advanced MCRs is performed by soft control. Using soft controls such as mouse control, touch screens, and so on, operators can select a specific screen, then choose the controller, and finally manipulate the devices. However, because of the different interfaces between soft control and hardwired conventional type control, different basic human error probabilities (BHEPs) should be considered in the Human Reliability Analysis (HRA) for advanced MCRs. Although there are many HRA methods to assess human reliabilities, such as Technique for Human Error Rate Prediction (THERP), Accident Sequence Evaluation Program (ASEP), Human Error Assessment and Reduction Technique (HEART), Human Event Repository and Analysis (HERA), Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR), Cognitive Reliability and Error Analysis Method (CREAM), and so on, these methods have been applied to conventional MCRs, and they do not consider the new features of advance MCRs such as soft controls. As a result, there is an insufficient database for assessing human reliabilities in advanced

  2. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  3. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  4. High-beta tokamak research. Annual progress report, 1 August 1982-1 August 1983

    International Nuclear Information System (INIS)

    The main research objectives during the past year fell into four areas: (1) detailed observations over a range of high-beta tokamak equilibria; (2) fabrication of an improved and more flexible high-beta tokamak based on our understanding of the present Torus II; (3) extension of the pulse length to 100 usec with power crowbar operation of the equilibrium field coil sets; and (4) comparison of our equilibrium and stability observations with computational models of MHD equilibrium and stability

  5. Development of the Fast Ionization Gauge in the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    WANGMingxu; LIBo; YANGZhigang; LIAOZhiqing; YANLongwen; ZHANGNianman; YANDonghai

    2003-01-01

    The neutral gas pressure near plasma or divertor plates is very important for the plasma-wall interaction, which determine the operation mode of divertom and confinement performances of plasma in tokamaks. The commercial ionization gauge does not work in strong magnetic field and noisy enviroment encountered in tokamaks. The measuring errom of pressure commercial ionizationare very large by the gauge mounted on the pumping system or through a long pipe to the vacuum vessel. A new ionization gauge,

  6. Digital controlled pulsed electric system of the ETE tokamak. First report

    International Nuclear Information System (INIS)

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author)

  7. Maintenance concept development for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs

  8. Magnetic diagnostics for the lithium tokamak experiment.

    Science.gov (United States)

    Berzak, L; Kaita, R; Kozub, T; Majeski, R; Zakharov, L

    2008-10-01

    The lithium tokamak experiment (LTX) is a spherical tokamak with R(0)=0.4 m, a=0.26 m, B(TF) approximately 3.4 kG, I(P) approximately 400 kA, and pulse length approximately 0.25 s. The focus of LTX is to investigate the novel low-recycling lithium wall operating regime for magnetically confined plasmas. This regime is reached by placing an in-vessel shell conformal to the plasma last closed flux surface. The shell is heated and then coated with liquid lithium. An extensive array of magnetic diagnostics is available to characterize the experiment, including 80 Mirnov coils (single and double axis, internal and external to the shell), 34 flux loops, 3 Rogowskii coils, and a diamagnetic loop. Diagnostics are specifically located to account for the presence of a secondary conducting surface and engineered to withstand both high temperatures and incidental contact with liquid lithium. The diagnostic set is therefore fabricated from robust materials with heat and lithium resistance and is designed for electrical isolation from the shell and to provide the data required for highly constrained equilibrium reconstructions. PMID:19044600

  9. Physics evaluation of compact tokamak ignition experiments

    International Nuclear Information System (INIS)

    At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/2/q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs

  10. System studies of compact ignition tokamaks

    International Nuclear Information System (INIS)

    The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B2a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations

  11. Boronization of Russian tokamaks from carborane precursors

    International Nuclear Information System (INIS)

    A new and cheap boronization technique using the nontoxic and nonexplosive solid substance carborane has been developed and successfully applied to the Russian tokamaks T-11M, T-3M, T-10 and TUMAN-3. The glow discharge in a mixture of He and carborane vapor produced the amorphous B/C coating with the B/C ratio varied from 2.0-3.7. The deposition rate was about 150 nm/h. The primary effect of boronization was a significant reduction of the impurity influx and the plasma impurity contamination, a sharp decrease of the plasma radiated power, and a decrease of the effective charge. Boronization strongly suppressed the impurity influx caused by additional plasma heating. ECR- and ICR-heating as well as ECR current drive were more effective in boronized vessels. Boronization resulted in a significant extension of the Ne- and q-region of stable tokamak operation. The density limit rose strongly. In Ohmic H-mode energy confinement time increased significantly (by a factor of 2) after boronization. It rose linearly with plasma current Ip and was 10 times higher than Neo-Alcator time at maximum current. ((orig.))

  12. System studies of compact ignition tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Galambos, J.D.; Blackfield, D.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Selcow, E.

    1987-08-01

    The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B/sup 2/a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations.

  13. Alfven wave studies on a tokamak

    International Nuclear Information System (INIS)

    The continuum modes of the shear Alfven resonance are studied on the Tokapole II device, a small tokamak operated in a four node poloidal divertor configuration. A variety of antenna designs and the efficiency with which they deliver energy to the resonant layer are discussed. The spatial structure of the driven waves is studied by means of magnetic probes inserted into the current channel. In an attempt to optimize the coupling of energy in to the resonant layer, the angle of antenna currents with respect to the equilibrium field, antenna size, and plasma-to-antenna distance are varied. The usefulness of Faraday shields, particle shields, and local limiters are investigated. Antennas should be well shielded, either a dense Faraday shield or particle shield being satisfactory. The antenna should be large and very near to the plasma. The wave magnetic fields measured show a spatial resonance, the position of which varies with the value of the equilibrium field and mass density. They are polarized perpendicular to the equilibrium field. A wave propagates radially in to the resonant surface where it is converted to the shear Alfven wave. The signal has a short risetime and does not propagate far toroidally. These points are all consistent with a strongly damped shear Alfven wave. Comparisons of this work to theoretical predictions and results from other tokamaks are made

  14. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    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.

  15. Design and construction of Alborz tokamak vacuum vessel system

    Energy Technology Data Exchange (ETDEWEB)

    Mardani, M., E-mail: mohsenmardani@gmail.com [Amirkabir University of Technology (Tehran Polytechnic), Tehran (Iran, Islamic Republic of); Amrollahi, R.; Koohestani, S. [Amirkabir University of Technology (Tehran Polytechnic), Tehran (Iran, Islamic Republic of)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. Black-Right-Pointing-Pointer As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. Black-Right-Pointing-Pointer 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. Black-Right-Pointing-Pointer 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.

  16. Dynamic diagnostics of the error fields in tokamaks

    Science.gov (United States)

    Pustovitov, V. D.

    2007-07-01

    The error field diagnostics based on magnetic measurements outside the plasma is discussed. The analysed methods rely on measuring the plasma dynamic response to the finite-amplitude external magnetic perturbations, which are the error fields and the pre-programmed probing pulses. Such pulses can be created by the coils designed for static error field correction and for stabilization of the resistive wall modes, the technique developed and applied in several tokamaks, including DIII-D and JET. Here analysis is based on the theory predictions for the resonant field amplification (RFA). To achieve the desired level of the error field correction in tokamaks, the diagnostics must be sensitive to signals of several Gauss. Therefore, part of the measurements should be performed near the plasma stability boundary, where the RFA effect is stronger. While the proximity to the marginal stability is important, the absolute values of plasma parameters are not. This means that the necessary measurements can be done in the diagnostic discharges with parameters below the nominal operating regimes, with the stability boundary intentionally lowered. The estimates for ITER are presented. The discussed diagnostics can be tested in dedicated experiments in existing tokamaks. The diagnostics can be considered as an extension of the 'active MHD spectroscopy' used recently in the DIII-D tokamak and the EXTRAP T2R reversed field pinch.

  17. Disruption avoidance through active magnetic feedback in tokamak plasmas

    Science.gov (United States)

    Paccagnella, Roberto; Zanca, Paolo; Yanovskiy, Vadim; Finotti, Claudio; Manduchi, Gabriele; Piron, Chiara; Carraro, Lorella; Franz, Paolo; RFX Team

    2014-10-01

    Disruptions avoidance and mitigation is a fundamental need for a fusion relevant tokamak. In this paper a new experimental approach for disruption avoidance using active magnetic feedback is presented. This scheme has been implemented and tested on the RFX-mod device operating as a circular tokamak. RFX-mod has a very complete system designed for active mode control that has been proved successful for the stabilization of the Resistive Wall Modes (RWMs). In particular the current driven 2/1 mode, unstable when the edge safety factor, qa, is around (or even less than) 2, has been shown to be fully and robustly stabilized. However, at values of qa (qa > 3), the control of the tearing 2/1 mode has been proved difficult. These results suggested the idea to prevent disruptions by suddenly lowering qa to values around 2 where the tearing 2/1 is converted to a RWM. Contrary to the universally accepted idea that the tokamaks should disrupt at low qa, we demonstrate that in presence of a well designed active control system, tokamak plasmas can be driven to low qa actively stabilized states avoiding plasma disruption with practically no loss of the plasma internal energy.

  18. Enterprise SRS: Leveraging Ongoing Operations to Advance Nuclear Fuel Cycle Programs - 12579

    International Nuclear Information System (INIS)

    on an individual sponsoring office. Given that reality, success for the current and future nuclear separations missions is dependent on a concerted effort to develop new, creative, approaches that leverage existing facilities in a manner that supports both near- and long-term needs of national programs. As a result of this situation, the Savannah River National Laboratory (SRNL) organized the 'Nuclear Separations User Facility Strategy Session' in Washington, D.C. on July 29, 2011. This workshop brought together key stakeholders from DOE and the private sector to develop a strategy for using engineering-scale nuclear materials processing facilities to advance our nation's nuclear separations research needs. In particular, the meeting focused on recommending how these engineering-scale demonstration facilities, like the Savannah River Site H-Canyon, can be connected with smaller 'bench-scale' research activities to form a seamless approach that integrates across the continuum of RD and D of advanced separations technologies. Coming out of this workshop, a new vision has been developed for a collaborative research facility model that centers on H-Canyon. Unique to this approach is the fact that H-Canyon will continue to accomplish DOE's critical nuclear material processing missions, while simultaneously serving as an RD and D resource for the scientific and technical portions of the nuclear separations community. This paper describes the planned operations for H-Canyon in FY2012 and beyond and discusses how these operations fit within the context of a collaborative research facility model and support the ongoing fuel cycle research and development programs of the DOE. (authors)

  19. Measurements of the subcriticality using advanced technique of shooting source during operation of NPP reactors

    Science.gov (United States)

    Lebedev, G. V.; Petrov, V. V.; Bobylyov, V. T.; Butov, R. I.; Zhukov, A. M.; Sladkov, A. A.

    2014-12-01

    According to the rules of nuclear safety, the measurements of the subcriticality of reactors should be carried out in the process of performing nuclear hazardous operations. An advanced technique of shooting source of neutrons is proposed to meet this requirement. As such a source, a pulsed neutron source (PNS) is used. In order to realize this technique, it is recommended to enable a PNS with a frequency of 1-20 Hz. The PNS is stopped after achieving a steady-state (on average) number of neutrons in the reactor volume. The change in the number of neutrons in the reactor volume is measured in time with an interval of discreteness of ˜0.1 s. The results of these measurements with the application of a system of point-kinetics equations are used in order to calculate the sought subcriticality. The basic idea of the proposed technique used to measure the subcriticality is elaborated in a series of experiments on the Kvant assembly. The conditions which should be implemented in order to obtain a positive result of measurements are formulated. A block diagram of the basic version of the experimental setup is presented, whose main element is a pulsed neutron generator.

  20. Electrical lysis: dynamics revisited and advances in On-chip operation.

    Science.gov (United States)

    Morshed, Bashir; Shams, Maitham; Mussivand, Tofy

    2013-01-01

    Electrical lysis (EL) is the process of breaking the cell membrane to expose the internal contents under an applied high electric field. Lysis is an important phenomenon for cellular analysis, medical treatment, and biofouling control. This paper aims to review, summarize, and analyze recent advancements on EL. Major databases including PubMed, Ei Engineering Village, IEEE Xplore, and Scholars Portal were searched using relevant keywords. More than 50 articles published in English since 1997 are cited in this article. EL has several key advantages compared to other lysis techniques such as chemical, mechanical, sonication, or laser, including rapid speed of operation, ability to control, miniaturization, low cost, and low power requirement. A variety of cell types have been investigated for including protoplasts, E. coli, yeasts, blood cells, and cancer cells. EL has been developed and applied for decontamination, cytology, genetics, single-cell analysis, cancer treatment, and other applications. On-chip EL is a promising technology for multiplexed automated implementation of cell-sample preparation and processing with micro- or nanoliter reagents.

  1. Advances in Remote Sensing of Agriculture: Context Description, Existing Operational Monitoring Systems and Major Information Needs

    Directory of Open Access Journals (Sweden)

    Clement Atzberger

    2013-02-01

    Full Text Available Many remote sensing applications are devoted to the agricultural sector. Representative case studies are presented in the special issue “Advances in Remote Sensing of Agriculture”. To complement the examples published within the special issue, a few main applications with regional to global focus were selected for this review, where remote sensing contributions are traditionally strong. The selected applications are put in the context of the global challenges the agricultural sector is facing: minimizing the environmental impact, while increasing production and productivity. Five different applications have been selected, which are illustrated and described: (1 biomass and yield estimation, (2 vegetation vigor and drought stress monitoring, (3 assessment of crop phenological development, (4 crop acreage estimation and cropland mapping and (5 mapping of disturbances and land use/land cover (LULC changes. Many other applications exist, such as precision agriculture and irrigation management (see other special issues of this journal, but were not included to keep the paper concise. The paper starts with an overview of the main agricultural challenges. This section is followed by a brief overview of existing operational monitoring systems. Finally, in the main part of the paper, the mentioned applications are described and illustrated. The review concludes with some key recommendations.

  2. Operating the Advanced Test Reactor in today's economic and regulatory environment

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory, is the US Department of Energy's largest and most versatile test reactor. Base programs at ATR are planned well into the 21st century. The ATR and support facilities along with an overview of current programs will be reviewed, but the main focus of the presentation will be on the impact that today's economic and regulatory concerns have had on the operation of this test reactor. Today's economic and regulatory concerns have demanded more work be completed at lower cost while increasing the margin of safety. By the beginning of the 1990 s, federal budgets for research generally and particularly for nuclear research had decreased dramatically. Many national needs continued to require testing in the ATR; but demanded lower cost, increased efficiency, improved performance, and an increased margin of safety. At the same time budgets were decreasing, there was an increase in regulatory compliance activity. The new standards imposed higher margins of safety. The new era of greater openness and higher safety standards complemented research demands to work safer, smarter and more efficiently. Several changes were made at the ATR to meet the demands of the sponsors and public. Such changes included some workforce reductions, securing additional program sponsors, upgrading some facilities, dismantling other facilities, and implementing new safety programs. (author)

  3. Description of the tasks of control room operators in German nuclear power plants and support possibilities by advanced computer systems

    International Nuclear Information System (INIS)

    In course of the development of nuclear power plants the instrumentation and control systems and the information in the control room have been increasing substantially. With this background it is described which operator tasks might be supported by advanced computer aid systems with main emphasis to safety related information and diagnose facilities. Nevertheless, some of this systems under development may be helpful for normal operation modes too. As far as possible recommendations for the realization and test of such systems are made. (orig.)

  4. Assembly of Aditya upgrade tokamak

    International Nuclear Information System (INIS)

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a tokamak with divertor configuration. At present the existing ADITYA tokamak has been dismantled up to bottom plinth on which the whole assembly of toroidal field (TF) coils and vacuum vessel rested. The major components of ADITYA machine includes 20 TF coils and its structural components, 9 Ohmic coils and its clamps, 4 BV coils and its clamps as well as their busbar connections, vacuum vessel and its supports and buckling cylinder, which are all being dismantled. The re-assembly of the ADITYA Upgrade tokamak started with installation and positioning of new buckling cylinder and central solenoid (TR1) coil. After that the inner sections of TF coils are placed following which in-situ winding, installation, positioning and support mounting of two pairs of new inner divertor coils have been carried out. After securing the TF coils with top I-beams the new torus shaped vacuum vessel with circular cross-section in 2 halves have been installed. The assembly of TF structural components such as top and bottom guiding wedges, driving wedges, top and bottom compression ring, inner and outer fish plates and top inverted triangle has been carried out in an appropriate sequence. The assembly of outer sections of TF coils along with the proper placements of top auxiliary TR and vertical field coils with proper alignment and positioning with the optical metrology instrument mainly completes the reassembly. Detailed re-assembly steps and challenges faced during re-assembly will be discussed in this paper. (author)

  5. Effect of energy and momentum conservation on fluid resonances for resonant magnetic perturbations in a tokamak

    International Nuclear Information System (INIS)

    In this paper, the impact of momentum and energy conservation of the collision operator in the kinetic description for Resonant Magnetic Perturbations (RMPs) in a tokamak is studied. The particle conserving differential collision operator of Ornstein-Uhlenbeck type is supplemented with integral parts such that energy and momentum are conserved. The application to RMP penetration in a tokamak shows that energy conservation in the electron collision operator is important for the quantitative description of plasma shielding effects at the resonant surface. On the other hand, momentum conservation in the ion collision operator does not significantly change the results

  6. Improved Disturbance Observer (DOB) Based Advanced Feedback Control for Optimal Operation of a Mineral Grinding Process%Improved Disturbance Observer (DOB) Based Advanced Feedback Control for Optimal Operation of a Mineral Grinding Process

    Institute of Scientific and Technical Information of China (English)

    周平; 向波; 柴天佑

    2012-01-01

    Advanced feedback control for optimal operation of mineral grinding process is usually based on the model predictive control (MPC) dynamic optimization. Since the MPC does not handle disturbances directly by controller design, it cannot achieve satisfactory effects in controlling complex grinding processes in the presence of strong disturbances and large uncertainties. In this paper, an improved disturbance observer (DOB) based MPC advanced feedback control is proposed to control the multivariable grinding operation. The improved DOB is based on the optimal achievable H 2 performance and can deal with disturbance observation for the nonminimum-phase delay systems. In this DOB-MPC advanced feedback control, the higher-level optimizer computes the optimal operation points by maximize the profit function and passes them to the MPC level. The MPC acts as a presetting controller and is employed to generate proper pre-setpoint for the lower-level basic feedback control system. The DOB acts as a compensator and improves the operation performance by dynamically compensating the setpoints for the basic control system according to the observed various disturbances and plant uncertainties. Several simulations are performed to demonstrate the proposed control method for grinding process operation.

  7. Teaching Advanced Operation of an iPod-Based Speech-Generating Device to Two Students with Autism Spectrum Disorders

    Science.gov (United States)

    Achmadi, Donna; Kagohara, Debora M.; van der Meer, Larah; O'Reilly, Mark F.; Lancioni, Giulio E.; Sutherland, Dean; Lang, Russell; Marschik, Peter B.; Green, Vanessa A.; Sigafoos, Jeff

    2012-01-01

    We evaluated a program for teaching two adolescents with autism spectrum disorders (ASD) to perform more advanced operations on an iPod-based speech-generating device (SGD). The effects of the teaching program were evaluated in a multiprobe multiple baseline across participants design that included two intervention phases. The first intervention…

  8. On the economic prospects of nuclear fusion with tokamaks

    Science.gov (United States)

    Pfirsch, D.; Schmitter, K. H.

    1987-12-01

    A method of cost and construction energy estimation for tokamak fusion power stations conforming to the present stage of fusion development is described. The method is based on first-wall heat load constraints rather than Beta limitations, which, however, might eventually be the more critical of the two. It is used to discuss the economic efficiency of pure fusion, with particular reference to the European study entitled Environmental Impact and Economic Prospects of Nuclear Fusion (1986). It is shown that the claims made therein for the economic prospects of pure fusion with tokamaks, when discussed on the basis of the present-day technology, do not stand up to critical examination. A fusion-fission hybrid, however, could afford more positive prospects. Support for the stated method is derived when it is properly applied for cost estimation of advanced gas-cooled and Magnox reactors, the two examples presented by the European study to disprove it.

  9. Analytical solutions for Tokamak equilibria with reversed toroidal current

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Caroline G. L.; Roberto, M.; Braga, F. L. [Departamento de Fisica, Instituto Tecnologico de Aeronautica, Sao Jose dos Campos, Sao Paulo 12228-900 (Brazil); Caldas, I. L. [Instituto de Fisica, Universidade de Sao Paulo, 05315-970 Sao Paulo, SP (Brazil)

    2011-08-15

    In tokamaks, an advanced plasma confinement regime has been investigated with a central hollow electric current with negative density which gives rise to non-nested magnetic surfaces. We present analytical solutions for the magnetohydrodynamic equilibria of this regime in terms of non-orthogonal toroidal polar coordinates. These solutions are obtained for large aspect ratio tokamaks and they are valid for any kind of reversed hollow current density profiles. The zero order solution of the poloidal magnetic flux function describes nested toroidal magnetic surfaces with a magnetic axis displaced due to the toroidal geometry. The first order correction introduces a poloidal field asymmetry and, consequently, magnetic islands arise around the zero order surface with null poloidal magnetic flux gradient. An analytic expression for the magnetic island width is deduced in terms of the equilibrium parameters. We give examples of the equilibrium plasma profiles and islands obtained for a class of current density profile.

  10. On the economic prospects of nuclear fusion with tokamaks

    International Nuclear Information System (INIS)

    This paper describes a method of cost and construction energy estimation for tokamak fusion power stations conforming to the present, early stage of fusion development. The method is based on first-wall heat load constraints rather than β limitations, which, however, might eventually be the more critical of the two. It is used to discuss the economic efficiency of pure fusion, with particular reference to the European study entitled 'Environmental Impact and Economic Prospects of Nuclear Fusion'. It is shown that the claims made therein for the economic prospects of pure fusion with tokamaks, when discussed on the basis of the present-day technology, do not stand up to critical examination. A fusion-fission hybrid, however, could afford more positive prospects. Support for the stated method is even derived when it is properly applied for cost estimation of advanced gascooled and Magnox reactors, the two very examples presented by the European study to 'disprove' it. (orig.)

  11. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Jae; Cheong, Y. H.; Choi, Y. S.; Cheon, E. J. [PlaGen, Seoul (Korea, Republic of)

    2003-11-15

    The establishment of regulatory and approval systems for KSTAR (Korea Superconducting Tokamak Advanced Research) has been demanded as the facility is targeted to be completed in the year of 2005. Such establishment can be achieved by performing adequate and in-depth analyses on safety issues covering radiological and chemical hazard materials, radiation protection, high vacuum, very low temperature, etc. The loss of coolant accidents and the loss of vacuum accident in fusion facilities have been introduced with summary of simulation results that were previously reported for ITER and JET. Computer codes that are actively used for accident simulation research are examined and their main features are briefly described. It can be stated that the safety analysis is indispensable to secure the safety of workers and individual members of the public as well as to establish the regulatory and approval systems for KSTAR tokamak.

  12. Recent experiments in the EAST and HT-7 superconducting tokamaks

    Science.gov (United States)

    Wan, Baonian; International EAST Collaborators; HT-7 Teams

    2009-10-01

    First divertor plasma configuration in Experimental Advanced Superconducting Tokamak (EAST) was obtained in the second campaign after the last IAEA meeting. To achieve long pulse diverted plasma discharges, new capabilities including the fully actively water cooled in-vessel components, current drive and heating systems, diagnostics and real-time plasma control algorithm were developed. Pre-programmed shape and feedback control of plasma position and current (RZIP) produced a variety of shaped plasma configurations, covering most of the configurations foreseen at the design stage of the machine. Control algorithm based on real-time equilibrium reconstruction and iso-flux control for the last closed magnetic flux surface (RTEFIT/ISOFLUX) has also been realized. A number of operational issues, such as plasma initiation and ramp up under constraints of superconducting coils were successfully investigated. First LHCD experiments demonstrated long pulse discharges longer than 20 s and nearly full non-inductive current drive. The physical engineering capability on the superconducting magnetic system was assessed by simulating discharges. Since the last IAEA meeting, experiments in HT-7 have been focusing on long pulse operation to support the EAST experiments on both physics and technical aspects. Long pulse discharges up to 400 s have now been achieved in HT-7. Investigation of sawtooth activities in ohmic and LHCD plasmas supports the turbulence model instead of the fast reconnection of the m = 1 magnetic island. Coexistence of electron mode and ion mode in high density ohmic plasmas has been observed by 2D ECE imaging (ECEI) in HT-7. The spectral characteristics of geodesic acoustic mode at the plasma boundary have been investigated by Langmuir probe arrays.

  13. Control-oriented Automatic System for Transport Analysis (ASTRA)-Matlab integration for Tokamaks

    International Nuclear Information System (INIS)

    The exponential growth in energy consumption has led to a renewed interest in the development of alternatives to fossil fuels. Between the unconventional resources that may help to meet this energy demand, nuclear fusion has arisen as a promising source, which has given way to an unprecedented interest in solving the different control problems existing in nuclear fusion reactors such as Tokamaks. The aim of this manuscript is to show how one of the most popular codes used to simulate the performance of Tokamaks, the Automatic System For Transport Analysis (ASTRA) code, can be integrated into the Matlab-Simulink tool in order to make easier and more comfortable the development of suitable controllers for Tokamaks. As a demonstrative case study to show the feasibility and the goodness of the proposed ASTRA-Matlab integration, a modified anti-windup Proportional Integral Derivative (PID)-based controller for the loop voltage of a Tokamak has been implemented. The integration achieved represents an original and innovative work in the Tokamak control area and it provides new possibilities for the development and application of advanced control schemes to the standardized and widely extended ASTRA transport code for Tokamaks. -- Highlights: → The paper presents a useful tool for rapid prototyping of different solutions to deal with the control problems arising in Tokamaks. → The proposed tool embeds the standardized Automatic System For Transport Analysis (ASTRA) code for Tokamaks within the well-known Matlab-Simulink software. → This allows testing and combining diverse control schemes in a unified way considering the ASTRA as the plant of the system. → A demonstrative Proportional Integral Derivative (PID)-based case study is provided to show the feasibility and capabilities of the proposed integration.

  14. Beam-induced tensor pressure tokamak equilibria

    International Nuclear Information System (INIS)

    D-shaped tensor pressure tokamak equilibria induced by neutral-beam injection are computed. The beam pressure components are evaluated from the moments of a distribution function that is a solution of the Fokker-Planck equation in which the pitch-angle scattering operator is ignored. The level-psub(perpendicular) contours undergo a significant shift away from the outer edge of the device with respect to the flux surfaces for perpendicular beam injection into broad-pressure-profile equilibria. The psub(parallel) contours undergo a somewhat smaller inward shift with respect to the flux surfaces for both parallel and perpendicular injection into broad-pressure-profile equilibria. For peaked-pressure-profile equilibria, the level pressure contours nearly co-incide with the flux surfaces. (author)

  15. Real time analysis of tokamak discharge parameters

    International Nuclear Information System (INIS)

    The techniques used in implementing two applications of real time analysis of data from the DIII-D tokamak are described. These tasks, which are demanding in both the speed of data acquisition and the speed of computation, execute on hardware capable of acquiring 40 million data samples per second and executing 80 million floating point operations per second. In the first case, a feedback control algorithm executing at a 10 kHz cycle frequency is used to specify the current in the poloidal field coils in order to control the discharge shape. In the second, fast Fourier transforms of Mirnov probe data are used to find the amplitude and frequency of each of eight toroidal mode numbers as a function of time during the discharge. Data sampled continuously at 500 kHz are used to produce results at 2 msec intervals

  16. Waste incineration models for operation optimization. Phase 1: Advanced measurement equipment for improved operation of waste fired plants; Affaldsforbraendingsmodeller til driftsoptimering. Fase 1: Avanceret maeleudstyr til forbedret drift af affaldsfyrede anlaeg

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-01

    This report describes results from the PSO projects ELTRA-5294 and ELTRA-5348: Waste incineration models for operation optimization. Phase 1, and Advanced measurement equipment for improved operation of waste fired plants. Phase 1. The two projects form the first step in a project course build on a long-term vision of a fully automatic system using a wide range of advanced measurement data, advanced dynamic models for prediction of operation and advanced regulation methods for optimization of the operation of waste incinerator plants. (BA)

  17. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    Science.gov (United States)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  18. Options for an ignited tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon ..beta../sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed.

  19. A compact Tokamak transmutation reactor

    Institute of Scientific and Technical Information of China (English)

    QiuLi-Jian; XiaoBing-Jia

    1997-01-01

    The low aspect ration tokamak is proposed for the driver of a transmutation reactor.The main parameters of the reactor core,neutronic analysis of the blanket are given>the neutron wall loading can be lowered from the magnitude order of 1 MW/m2 to 0.5MW/m2 which is much easier to reach in the near future,and the transmutation efficiency (fission/absorption ratio)is raised further.The blanket power density is about 200MW/m3 which is not difficult to deal with.The key components such as diverter and center conductor post are also designed and compared with conventional TOkamak,Finally,by comparison with the other drivers such as FBR,PWR and accelerator,it can be anticipated that the low aspect ratio transmutation reactor would be one way of fusion energy applications in the near future.

  20. Equilibrium Reconstruction in EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    QIAN Jinping; WAN Baonian; L. L. LAO; SHEN Biao; S. A. SABBAGH; SUN Youwen; LIU Dongmei; XIAO Singjia; REN Qilong; GONG Xianzu; LI Jiangang

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of toka-mak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier ex-pansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.

  1. First 3 years of operation of RIACS (Research Institute for Advanced Computer Science) (1983-1985)

    Science.gov (United States)

    Denning, P. J.

    1986-01-01

    The focus of the Research Institute for Advanced Computer Science (RIACS) is to explore matches between advanced computing architectures and the processes of scientific research. An architecture evaluation of the MIT static dataflow machine, specification of a graphical language for expressing distributed computations, and specification of an expert system for aiding in grid generation for two-dimensional flow problems was initiated. Research projects for 1984 and 1985 are summarized.

  2. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb3Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered

  3. Commissioning of heating neutral beams for COMPASS-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Deichuli, P.; Davydenko, V.; Belov, V.; Gorbovsky, A.; Dranichnikov, A.; Ivanov, A.; Sorokin, A.; Mishagin, V.; Abdrashitov, A.; Kolmogorov, V.; Kondakov, A. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation)

    2012-02-15

    Two neutral beam injectors have been developed for plasma heating on COMPASS-D tokamak (Institute of Plasma Physics, Prague). The 4-electrodes multihole ion-optical system with beam focusing was chosen to provide the low divergence 300 kW power in both deuterium and hydrogen atoms. The accelerating voltage is 40 kV at extracted ion current up to 15 A. The power supply system provides the continuous and modulated mode of the beam injection at a maximal pulse length 300 ms. The optimal arrangement of the cryopanels and the beam duct elements provides sufficiently short-length beamline which reduces the beam losses. The evolution of the impurities and molecular fraction content is studied in the process of the high voltage conditioning of the newly made ion sources. Two injectors of the same type have been successfully tested and are ready for operation at tokamak in IPP, Prague.

  4. Instrumentation for plasma diagnosis in TN (Novillo Tokamak)

    International Nuclear Information System (INIS)

    In the Plasma Physics Laboratory of National Institute of Nuclear Research it has been utilized different devices for to determine electromagnetic parameters of Novillo Tokamak such as: magnetic fields, plasma currents, plasma column position and hoop voltage. For these measurements it was designed, constructed and calibrated magnetic soundings such as: magnetic field soundings, Rogowsky coil, coils of the type called sine/cosine and spires type riding saddle; as well as the electronic instrumentation associated with these devices. This electronics to be clear of instrumentation amplifiers for the detection of the soundings signals and differentiators utilized for the elimination of spurious induced currents in the soundings by the different Novillo electromagnetic fields. In this work is presented the methodology for the construction of this instruments, as well as the results of measurements effectuated in the two operation regimens of Tokamak: Cleaning discharge and Main discharge. (Author)

  5. A quasi-linear gyrokinetic transport model for tokamak plasmas

    CERN Document Server

    Casati, Alessandro

    2012-01-01

    The development of a quasi-linear gyrokinetic transport model for tokamak plasmas, ultimately designed to provide physically comprehensive predictions of the time evolution of the thermodynamic relevant quantities, is a task that requires tight links among theoretical, experimental and numerical studies. The framework of the model here proposed, which operates a reduction of complexity on the nonlinear self-organizing plasma dynamics, allows in fact multiple validations of the current understanding of the tokamak micro-turbulence. The main outcomes of this work stem from the fundamental steps involved by the formulation of such a reduced transport model, namely: (1) the verification of the quasi-linear plasma response against the nonlinearly computed solution, (2) the improvement of the turbulent saturation model through an accurate validation of the nonlinear codes against the turbulence measurements, (3) the integration of the quasi-linear model within an integrated transport solver.

  6. DIII-D RESEARCH OPERATIONS ANNUAL REPORT October 1, 2001 through September 30, 2002

    International Nuclear Information System (INIS)

    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

  7. Development of advanced inductive scenarios for ITER

    Science.gov (United States)

    Luce, T. C.; Challis, C. D.; Ide, S.; Joffrin, E.; Kamada, Y.; Politzer, P. A.; Schweinzer, J.; Sips, A. C. C.; Stober, J.; Giruzzi, G.; Kessel, C. E.; Murakami, M.; Na, Y.-S.; Park, J. M.; Polevoi, A. R.; Budny, R. V.; Citrin, J.; Garcia, J.; Hayashi, N.; Hobirk, J.; Hudson, B. F.; Imbeaux, F.; Isayama, A.; McDonald, D. C.; Nakano, T.; Oyama, N.; Parail, V. V.; Petrie, T. W.; Petty, C. C.; Suzuki, T.; Wade, M. R.; the ITPA Integrated Operation Scenario Topical Group Members; the ASDEX-Upgrade Team; the DIII-D Team; EFDA Contributors, JET; the JT-60U Team

    2014-01-01

    Since its inception in 2002, the International Tokamak Physics Activity topical group on Integrated Operational Scenarios (IOS) has coordinated experimental and modelling activity on the development of advanced inductive scenarios for applications in the ITER tokamak. The physics basis and the prospects for applications in ITER have been advanced significantly during that time, especially with respect to experimental results. The principal findings of this research activity are as follows. Inductive scenarios capable of higher normalized pressure (βN ⩾ 2.4) than the ITER baseline scenario (βN = 1.8) with normalized confinement at or above the standard H-mode scaling are well established under stationary conditions on the four largest diverted tokamaks (AUG, DIII-D, JET, JT-60U), demonstrated in a database of more than 500 plasmas from these tokamaks analysed here. The parameter range where high performance is achieved is broad in q95 and density normalized to the empirical density limit. MHD modes can play a key role in reaching stationary high performance, but also define the limits to achieved stability and confinement. Projection of performance in ITER from existing experiments uses empirical scalings and theory-based modelling. The status of the experimental validation of both approaches is summarized here. The database shows significant variation in the energy confinement normalized to standard H-mode confinement scalings, indicating the possible influence of additional physics variables absent from the scalings. Tests using the available information on rotation and the ratio of the electron and ion temperatures indicate neither of these variables in isolation can explain the variation in normalized confinement observed. Trends in the normalized confinement with the two dimensionless parameters that vary most from present-day experiments to ITER, gyroradius and collision frequency, are significant. Regression analysis on the multi-tokamak database has been

  8. High performance discharges in the Lithium Tokamak eXperiment with liquid lithium wallsa)

    Science.gov (United States)

    Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G.; Capece, A.; Koel, B.; Roszell, J.; Biewer, T. M.; Gray, T. K.; Kubota, S.; Beiersdorfer, P.; Widmann, K.; Tritz, K.

    2015-05-01

    The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

  9. Langmuir-magnetic probe measurements of ELMs and dithering cycles in the EAST tokamak

    DEFF Research Database (Denmark)

    Yan, Ning; Naulin, Volker; Xu, G. S.;

    2014-01-01

    Measurements of the dynamical behavior associated with edge localized modes (ELMs) have been carried out in the Experimental Advanced Superconducting Tokamak (EAST) by direct probing near the separatrix and far scrape-off layer (SOL) using electrostatic as well as magnetic probes. Type-III ELMs a...

  10. A comparison of steady-state ARIES and pulsed PULSAR tokamak power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.

    1994-07-01

    The multi-institutional ARIES study has completed a series of three steady-state and two pulsed cost-optimized conceptual designs of commercial tokamak fusion power plants that vary the level of assumed advances in technology and physics. The cost benefits of various design options are compared quantitatively. Possible means to improve the economic competitiveness of fusion are suggested.

  11. Preliminary Design of the Data Processing Program on the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    YANGYang; CHENLiaoyuan; PANYudong; PANLi

    2002-01-01

    In this paper, by means of the two program tools, C+ + Builder and Matlab, the preliminary design of the data processing for experiment on the HL-2A tokamak is described. The software of data processing includes two parts: common processing software and advanced processing software.

  12. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form

  13. The disruptive instability in Tokamak plasmas

    NARCIS (Netherlands)

    Salzedas, F.J.B.

    2001-01-01

    Studies performed in RTP (Rijnhuizen Tokamak Project) of the most violent and dangerous instability in tokamak plasmas, the major disruption, are presented. A particular class of disruptions is analyzed, namely the density limit disruption, which occur in high density plasmas. The radiative te

  14. The role of limiter in Egyptor Tokamak

    CERN Document Server

    Ei-Sisi, A B

    2002-01-01

    In Egyptor Tokamak, the limiter is used for separation of the plasma from the vessel. In this work an overview of limiter types, and construction of limiter in Egyptor Tokamak is discussed. Also simulation results of the radial electron density distribution in case of limiter are presented. The results of the simulation are in agreement with the experimental and analytical results.

  15. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-07-01

    Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The

  16. Experimental results from the TFTR tokamak

    International Nuclear Information System (INIS)

    Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam-heated discharges. The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2 T). Initial neutral-beam-heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 1020 m-3 in high current discharges. At the opposite extreme, by operating at low plasma current (I/sub p/ ∼ 0.8 MA) and low density (anti n/sub e/ ∼ 1 x 1019 m-3), high ion temperatures (9 +- 2 keV) and rotation speeds (7 x 105 m/s) have been achieved during injection. In addition, plasma compression experiments have demonstrated acceleration of beam ions from 82 keV to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam-heated discharges. The results of these confinement studies are presented

  17. Deposit of thin films for Tokamaks conditioning; Deposito de peliculas delgadas para acondicionar Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Valencia A, R

    2006-07-01

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (<10 eV), both in noble and reactive gases, as well as the conditioning by thin film deposits of hydrogen rich amorphous carbon (carbonization) leading to a reduction in the plasma resistivity from 8.99 x 10{sup -6} to 4.5 x 10{sup -6} {omega}-m, thus taking the Z{sub ef} value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission

  18. KTM Tokamak is prototype of X XI century reactor. Future International laboratory of thermonuclear materials testing and power engineering

    International Nuclear Information System (INIS)

    In 29-31 May of 2000 the presentation of the joint Kazakhstan-Russian draft of Kazakhstan material-testing tokamak (KTM) was carried out. KTM tokamak is implementing by decision of the President and Government of Republic of Kazakhstan for supporting of the Kazakhstan participation in development of draft within framework of ITER fusion reactor construction. Scientific head of the project is Russian academician - Velikhov V. and Russian Research Center 'Kurchatovskij Institute' , General designers - Scientific Research Institute for Electrophysical Equipment after D. V. Efremov (Russian Federation) and Kazakh Research Inst. for Energy Industry (KazNIIEhnergoprom). Scientific part of the project is working out in National Nuclear Center of Republic of Kazakhstan and Scientific Research Institute of Experimental and Theoretical Physics. KTM tokamak is experimental fusion device for materials testing study, as well as for designing of methods for protection of the reactor first wall, in-chamber elements and divertor planes, high frequency heat of antennas in energetic load regimes close to both the ITER and the future fusion energy reactors. KTM by it design presents spheric tokamak, which successfully combining advantages of the spheromaks (compactness) and the tokamaks (high plasma density). Now in the world there are similar operating spheric tokamaks: NSTX (USA), MAST (Great Britain), GLOBUS-M (Russian Federation). Principal peculiarity of KTM tokamak is existence of moving divertor device, which with help of manipulator allows to changing of examining samples without high vacuum disruption. Values of the thermal loads and fluences in the KTM are equal or higher than loads in operating tokamaks and correspond with ITER reactor loads. KTM tokamak will be the only mega-ampere device in the world with the aspect ratio A=2

  19. 20 years of research on the Alcator C-Mod tokamak

    OpenAIRE

    Greenwald, Martin; Bader, A; Baek, S.; M. Bakhtiari; Barnard, H.; Beck, W.; Bergerson, W; Bespamyatnov, I; Bonoli, P.; Brower, D; Brunner, D.; Burke, W.; Candy, J.; Churchill, M; Cziegler, I.

    2014-01-01

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of crit...

  20. Equilibrium reconstruction based on core magnetic measurement and its applications on equilibrium transition in Joint-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Jian, X.; Li, Q.; Liu, Y.; Gao, L.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-10-15

    Evaluation and reconstruction of plasma equilibrium, especially to resolve the safety factor profile, is imperative for advanced tokamak operation and physics study. Based on core magnetic measurement by the high resolution laser polarimeter-interferometer system (POLARIS), the equilibrium of Joint-TEXT (J-TEXT) plasma is reconstructed and profiles of safety factor, current density, and electron density are, therefore, obtained with high accuracy and temporal resolution. The equilibrium reconstruction procedure determines the equilibrium flux surfaces essentially from the data of POLARIS. Refraction of laser probe beam, a major error source of the reconstruction, has been considered and corrected, which leads to improvement of accuracy more than 10%. The error of reconstruction has been systematically assessed with consideration of realistic diagnostic performance and scrape-off layer region of plasma, and its accuracy has been verified. Fast equilibrium transitions both within a single sawtooth cycle and during the penetration of resonant magnetic perturbation have been investigated.

  1. Technology Alignment and Portfolio Prioritization (TAPP): Advanced Methods in Strategic Analysis, Technology Forecasting and Long Term Planning for Human Exploration and Operations, Advanced Exploration Systems and Advanced Concepts

    Science.gov (United States)

    Funaro, Gregory V.; Alexander, Reginald A.

    2015-01-01

    The Advanced Concepts Office (ACO) at NASA, Marshall Space Flight Center is expanding its current technology assessment methodologies. ACO is developing a framework called TAPP that uses a variety of methods, such as association mining and rule learning from data mining, structure development using a Technological Innovation System (TIS), and social network modeling to measure structural relationships. The role of ACO is to 1) produce a broad spectrum of ideas and alternatives for a variety of NASA's missions, 2) determine mission architecture feasibility and appropriateness to NASA's strategic plans, and 3) define a project in enough detail to establish an initial baseline capable of meeting mission objectives ACO's role supports the decision­-making process associated with the maturation of concepts for traveling through, living in, and understanding space. ACO performs concept studies and technology assessments to determine the degree of alignment between mission objectives and new technologies. The first step in technology assessment is to identify the current technology maturity in terms of a technology readiness level (TRL). The second step is to determine the difficulty associated with advancing a technology from one state to the next state. NASA has used TRLs since 1970 and ACO formalized them in 1995. The DoD, ESA, Oil & Gas, and DoE have adopted TRLs as a means to assess technology maturity. However, "with the emergence of more complex systems and system of systems, it has been increasingly recognized that TRL assessments have limitations, especially when considering [the] integration of complex systems." When performing the second step in a technology assessment, NASA requires that an Advancement Degree of Difficulty (AD2) method be utilized. NASA has used and developed or used a variety of methods to perform this step: Expert Opinion or Delphi Approach, Value Engineering or Value Stream, Analytical Hierarchy Process (AHP), Technique for the Order of

  2. Operational transparency: an advanced safeguards strategy for future on-load refuelled reactors

    International Nuclear Information System (INIS)

    The IAEA's system for tracking fuel movement in an on-load refuelled heavy-water reactor is robust, but an opportunity remains to exploit the wealth of data streaming from the reactor vault during operation and provide real-time, third-party monitoring of reactor status and history. This concept of Operational Transparency would require that large amounts of operational data be reduced in near-real time to a small subset of high-level information. Operational Transparency would enhance the IAEA's ability to monitor the state of the core to an unprecedented level. This paper provides an overview of the novel concept of Operational Transparency in heavy water reactors, using potential application to CANDU reactors as an example, and explores some of the technical challenges that will need to be solved for efficient implementation. (author)

  3. Advanced Tele-operation[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M.

    1998-07-01

    Maintenance, repair, and dismantling operations in nuclear facilities have to be performed remotely when high radiation doses exclude hands-on operation, but also to minimize contamination risks and occupational doses to the operators. Computer-aided and sensor-based tele-operation enhances safety, reliability, and performance by helping the operator in difficult tasks with poor remote environmental perception. The objectives of work in this domain are to increase the scientific knowledge of the studied phenomena, to improve the interpretation of data, to improve the piloting og experimental devices during irradiation, to reveal and to understand possible unexpected phenomena occurring during irradiation. This scientific report describes the achievements for 1997 in the area of radiation tolerance for of remote-sensing, optical fibres and optical fibre sensors.

  4. Review of the Equilibrium Fitting for Non-Circular Tokamak

    Institute of Scientific and Technical Information of China (English)

    罗家融

    2002-01-01

    As the equilibrium fitting code (EFIT) is developing to perform the magnetic and the kinetic-magnetic analysis for tokamak device operation, it can be not only run in either the fitting mode or the equilibrium mode but also control operation of modern experimental fusion device. In this paper the history of EF1T code and its capabilities are described in section 2. A brief description of the off-line EFIT code and the development of the real-time EFIT (RTEFIT)code is shown in section 3 and 4 respectively. In the last section the summary of this paper is given.

  5. Tokamak-7 cryogenic system and its test data

    International Nuclear Information System (INIS)

    The composition, structure and purpose of the Tokamak-7 device cryogenic system are described. The cryogenic system comprises the helium system and nitrogen cryogenic system. Heat inflows at different temperature levels to cryogenic objects under various modes of the device operation are estimated. The results of the cryogenic system autonomous tests under cooling, cryostating and heating of the T-7 device cryogenic objects are presented. The three year experience of cryogenic system operation for the circulation superconducting magnetic system is accumulated and analysed. Important results necessary for designing cryogenic complexes of newly developed large thremonuclear devices are obtained

  6. LONG-PULSE, HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

    Energy Technology Data Exchange (ETDEWEB)

    T.C. LUCE; M.R. WADE; P.A. POLITZER; S.L. ALLEN; M E. AUSTIN; D.R. BAKER; B.D. BRAY; D.P. BRENNAN; K.H. BURRELL; T.A. CASPER; M.S. CHU; J.D. De BOO; E.J. DOYLE; J.R. FERRON; A.M. GAROFALO; P.GOHIL; I.A. GORELOV; C.M. GREENFIELD; R.J. GROEBNER; W.W. HEIBRINK; C.-L. HSIEH; A.W. HYATT; R.JAYAKUMAR; J.E.KINSEY; R.J. LA HAYE; L.L.LAO; C.J.LASNIER; E.A. LAZARUS; A.W. LEONARD; Y.R.LIN-LIU; J.LOHR; M.A. MAKOWSKI; M.MURAKAMI; C.C.PETTY; R.I. PINSKER; R.PRATER; C.L. RETTIG; T.L. RHODES; B.W. RICE; E.J. STRAIT; T.S. TAYLOR; D.M. THOMAS; A.D. TURNBULL; J.G. WATKINS; W.P.WEST; K.-L. WONG

    2000-10-01

    Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance {approx}10 has been sustained for >5 {tau}{sub E} with q{sub min} >1.5. (The normalized performance is measured by the product {beta}{sub N} H{sub 89} indicating the proximity to the conventional {beta} limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and {beta} {approx}{le} 5%. The limit to increasing {beta} is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate {approx}75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and {beta} control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with {beta}{sub N}H{sub 89} {approx} 7 for up to 6.3 s or {approx} 34 {tau}{sub E}. These discharges appear to be in resistive equilibrium with q{sub min} {approx} 1.05, in agreement with the current profile relaxation time of 1.8 s.

  7. Breakdown in the pretext tokamak

    International Nuclear Information System (INIS)

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges

  8. The procedure execution manager and its application to Advanced Photon Source operation

    International Nuclear Information System (INIS)

    The Procedure Execution Manager (PEM) combines a complete scripting environment for coding accelerator operation procedures with a manager application for executing and monitoring the procedures. PEM is based on Tcl/Tk, a supporting widget library, and the dp-tcl extension for distributed processing. The scripting environment provides support for distributed, parallel execution of procedures along with join and abort operations. Nesting of procedures is supported, permitting the same code to run as a top-level procedure under operator control or as a subroutine under control of another procedure. The manager application allows an operator to execute one or more procedures in automatic, semi-automatic, or manual modes. It also provides a standard way for operators to interact with procedures. A number of successful applications of PEM to accelerator operations have been made to date. These include start-up, shutdown, and other control of the positron accumulator ring (PAR), low-energy transport (LET) lines, and the booster rf systems. The PAR/LET procedures make nested use of PEM's ability to run parallel procedures. There are also a number of procedures to guide and assist tune-up operations, to make accelerator physics measurements, and to diagnose equipment. Because of the success of the existing procedures, expanded use of PEM is planned

  9. Experience and technical issues of liquid lithium application as plasma facing material in tokamaks

    International Nuclear Information System (INIS)

    The following critical issues of liquid lithium used in tokamak conditions are considered: major physical properties of lithium, physico-chemical aspects of lithium interaction and compatibility with structural materials of fusion reactors. Lithium capillary-porous system (CPS) is considered as advanced plasma facing material for power fusion reactor and its main properties are presented. Review of plasma facing element (PFE) structures based on lithium CPS and tests results in T-11M, T-10 and FTU tokamaks are included. Brief review of projects of lithium limiter of FTU with active system for thermal stabilization and module of lithium divertor for KTM tokamak with liquid metal (Na-K) cooling system based on the lithium CPS use are presented.

  10. Experimental study of external kink instabilities in the Columbia High Beta Tokamak

    International Nuclear Information System (INIS)

    The generation of power through controlled thermonuclear fusion reactions in a magnetically confined plasma holds promise as a means of supplying mankind's future energy needs. The device most technologically advanced in pursuit of this goal is the tokamak, a machine in which a current-carrying toroidal plasma is thermally isolated from its surroundings by a strong magnetic field. To be viable, the tokamak reactor must produce a sufficiently large amount of power relative to that needed to sustain the fusion reactions. Plasma instabilities may severely limit this possibility. In this work, I describe experimental measurements of the magnetic structure of large-scale, rapidly-growing instabilities that occur in a tokamak when the current or pressure of the plasma exceeds a critical value relative to the magnetic field, and I compare these measurements with theoretical predictions

  11. Advance of Hazardous Operation Robot and its Application in Special Equipment Accident Rescue

    Science.gov (United States)

    Zeng, Qin-Da; Zhou, Wei; Zheng, Geng-Feng

    A survey of hazardous operation robot is given out in this article. Firstly, the latest researches such as nuclear industry robot, fire-fighting robot and explosive-handling robot are shown. Secondly, existing key technologies and their shortcomings are summarized, including moving mechanism, control system, perceptive technology and power technology. Thirdly, the trend of hazardous operation robot is predicted according to current situation. Finally, characteristics and hazards of special equipment accident, as well as feasibility of hazardous operation robot in the area of special equipment accident rescue are analyzed.

  12. Comparative study of runaway electron diffusion in the rise phase of low and normal discharges in the SINP tokamak

    Indian Academy of Sciences (India)

    Ramesh Narayanan; A N Sekar Iyengar

    2010-10-01

    The behaviour of runaway electrons in the SINP tokamak, which can be operated in a normal edge safety factor () (NQ) discharge configuration as well as in a low (LQ) configuration, was experimentally investigated, during the initial plasma generation phase. An energy analysis of the runaway electron dynamics in the rise phase of the SINP tokamak discharges was also made. A comparison of the runaway electron diffusion coefficients in NQ and LQ is carried out in this paper.

  13. Dust-Particle Transport in Tokamak Edge Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K; Rognlien, T D

    2005-09-12

    Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.

  14. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)

  15. Performance Projections For The Lithium Tokamak Experiment (LTX)

    Energy Technology Data Exchange (ETDEWEB)

    Majeski, R.; Berzak, L.; Gray, T.; Kaita, R.; Kozub, T.; Levinton, F.; Lundberg, D. P.; Manickam, J.; Pereverzev, G. V.; Snieckus, K.; Soukhanovskii, V.; Spaleta, J.; Stotler, D.; Strickler, T.; Timberlake, J.; Yoo, J.; Zakharov, L.

    2009-06-17

    Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized.

  16. Prototype Development of Remote Operated Hot Uniaxial Press (ROHUP) to Fabricate Advanced Tc-99 Bearing Ceramic Waste Forms - 13381

    Energy Technology Data Exchange (ETDEWEB)

    Alaniz, Ariana J.; Delgado, Luc R.; Werbick, Brett M. [University of Nevada - Las Vegas, Howard R. Hughes College of Engineering, 4505 S. Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States); Hartmann, Thomas [University of Nevada - Las Vegas, Harry Reid Canter, 4505 S. Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States)

    2013-07-01

    The objective of this senior student project is to design and build a prototype construction of a machine that simultaneously provides the proper pressure and temperature parameters to sinter ceramic powders in-situ to create pellets of rather high densities of above 90% (theoretical). This ROHUP (Remote Operated Hot Uniaxial Press) device is designed specifically to fabricate advanced ceramic Tc-99 bearing waste forms and therefore radiological barriers have been included in the system. The HUP features electronic control and feedback systems to set and monitor pressure, load, and temperature parameters. This device operates wirelessly via portable computer using Bluetooth{sup R} technology. The HUP device is designed to fit in a standard atmosphere controlled glove box to further allow sintering under inert conditions (e.g. under Ar, He, N{sub 2}). This will further allow utilizing this HUP for other potential applications, including radioactive samples, novel ceramic waste forms, advanced oxide fuels, air-sensitive samples, metallic systems, advanced powder metallurgy, diffusion experiments and more. (authors)

  17. Advanced transport operating system software upgrade: Flight management/flight controls software description

    Science.gov (United States)

    Clinedinst, Winston C.; Debure, Kelly R.; Dickson, Richard W.; Heaphy, William J.; Parks, Mark A.; Slominski, Christopher J.; Wolverton, David A.

    1988-01-01

    The Flight Management/Flight Controls (FM/FC) software for the Norden 2 (PDP-11/70M) computer installed on the NASA 737 aircraft is described. The software computes the navigation position estimates, guidance commands, those commands to be issued to the control surfaces to direct the aircraft in flight based on the modes selected on the Advanced Guidance Control System (AGSC) mode panel, and the flight path selected via the Navigation Control/Display Unit (NCDU).

  18. Interactive exploration of tokamak turbulence simulations in virtual reality

    International Nuclear Information System (INIS)

    We have developed an immersive visualization system designed for interactive data exploration as an integral part of our computing environment for studying tokamak turbulence. This system of codes can reproduce the results of simulations visually for scrutiny in real time, interactively and with more realism than ever before. At peak performance, the VR system can present for view some 400 coordinated images per second. The long term vision this approach targets is a open-quote holodeck-like close-quote virtual-reality environment in which one can explore gyrofluid or gyrokinetic plasma simulations interactively and in real time, visually, with concurrent simulations of experimental diagnostic devices. In principle, such a open-quote virtual tokamak close-quote computed environment could be as all encompassing or as focussed as one likes, in terms of the physics involved. The computing framework in one within which a group of researchers can work together to produce a real and identifiable product with easy access to all contributions. This could be our version of NASA's next generation Numerical Wind Tunnel. The principal purpose of this VR capability for Numerical Tokamak simulation is to provide interactive visual experience to help create new ways of understanding aspects of the convective transport processes operating in tokamak fusion experiments. The effectiveness of the visualization method is strongly dependent on the density of frame-to-frame correlation. Below a threshold of this quantity, short term visual memory does not bridge the gap between frames well enough for there to exist a strong visual connection. Above the threshold, evolving structures appear clearly. The visualizations show the 3D structure of vortex evolution and the gyrofluid motion associated with it. We discovered that it was very helpful for visualizing the cross field flows to compress the virtual world in the toroidal angle

  19. Advances in the operational safety of nuclear power plants. Proceedings of an international symposium

    International Nuclear Information System (INIS)

    The main purpose of the Conference was to provide a forum for exchange of information among around 200 attending experts from 46 Member States and five international organizations, who altogether presented around 80 papers and posters. The Conference presentations were divided into four main topics: Managing and Regulating Safe Operation; Safety Performance and Lessons Learned; Improving Operational Safety Using PSA; Enhancing Safety. Refs, figs, tabs

  20. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m2 and a particle heat flux of 1 MW/m2. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  1. Atomic physics in tokamak plasmas

    International Nuclear Information System (INIS)

    Tokamak discharges produce hydrogen-isotope plasmas in a quasi-steady state, with radial electron temperature, Tsub(e)(r), and density nsub(e)(r), distribution usually centrally peaked, with typical values Tsub(e)(0) approx.= 1 - 3 keV, nsub(e)(r) approx.= 1014 cm-3. Besides hydrogen, the plasma contains small quantities of carbon, oxygen, various construction or wall-conditioning materials such as Fe, Cr, Ni, Ti, Zr, Mo, and perhaps elements added for special diagnostic purposes, e.g., Si, Sc, Al, or noble gases. These elements are spatially fairly homogeneously distributed, with the different ionization states occurring near radial locations where Tsub(e)(r) approx.= Esub(i), the ionization potential. Thus, spectroscopic measurements of various plasma properties, such as ion temperatures, plasma motions or oscillations, radial transport rates, etc. are automatically endowed with spatial resolution. Furthermore the emitted spectra, even of heavier elements such as Fe or Ni, are fairly simple because only the ground levels are appreciably populated under the prevailing plasma conditions. Identification of near-ground transitions, including particularly magnetic dipole and intercombination transitions of ions with ionization potentials in the several keV range, and determination of their collisional and radiative transition probabilities will be required for development of appropriate diagnostics of tokamak-type plasma approaching the prospective fusion reactor conditions. (orig.)

  2. Advanced thermal barrier coatings for operation in high hydrogen content fueled gas turbines.

    Energy Technology Data Exchange (ETDEWEB)

    Sampath, Sanjay [Stony Brook Univ., NY (United States)

    2015-04-02

    The Center for Thermal Spray Research (CTSR) at Stony Brook University in partnership with its industrial Consortium for Thermal Spray Technology is investigating science and technology related to advanced metallic alloy bond coats and ceramic thermal barrier coatings for applications in the hot section of gasified coal-based high hydrogen turbine power systems. In conjunction with our OEM partners (GE and Siemens) and through strategic partnership with Oak Ridge National Laboratory (ORNL) (materials degradation group and high temperature materials laboratory), a systems approach, considering all components of the TBC (multilayer ceramic top coat, metallic bond coat & superalloy substrate) is being taken during multi-layered coating design, process development and subsequent environmental testing. Recent advances in process science and advanced in situ thermal spray coating property measurement enabled within CTSR has been incorporated for full-field enhancement of coating and process reliability. The development of bond coat processing during this program explored various aspects of processing and microstructure and linked them to performance. The determination of the bond coat material was carried out during the initial stages of the program. Based on tests conducted both at Stony Brook University as well as those carried out at ORNL it was determined that the NiCoCrAlYHfSi (Amdry) bond coats had considerable benefits over NiCoCrAlY bond coats. Since the studies were also conducted at different cycling frequencies, thereby addressing an associated need for performance under different loading conditions, the Amdry bond coat was selected as the material of choice going forward in the program. With initial investigations focused on the fabrication of HVOF bond coats and the performance of TBC under furnace cycle tests , several processing strategies were developed. Two-layered HVOF bond coats were developed to render optimal balance of density and surface roughness

  3. Optimization of Sinter Plant Operating Conditions Using Advanced Multivariate Statistics: Intelligent Data Processing

    Science.gov (United States)

    Fernández-González, Daniel; Martín-Duarte, Ramón; Ruiz-Bustinza, Íñigo; Mochón, Javier; González-Gasca, Carmen; Verdeja, Luis Felipe

    2016-08-01

    Blast furnace operators expect to get sinter with homogenous and regular properties (chemical and mechanical), necessary to ensure regular blast furnace operation. Blends for sintering also include several iron by-products and other wastes that are obtained in different processes inside the steelworks. Due to their source, the availability of such materials is not always consistent, but their total production should be consumed in the sintering process, to both save money and recycle wastes. The main scope of this paper is to obtain the least expensive iron ore blend for the sintering process, which will provide suitable chemical and mechanical features for the homogeneous and regular operation of the blast furnace. The systematic use of statistical tools was employed to analyze historical data, including linear and partial correlations applied to the data and fuzzy clustering based on the Sugeno Fuzzy Inference System to establish relationships among the available variables.

  4. Optimization of Sinter Plant Operating Conditions Using Advanced Multivariate Statistics: Intelligent Data Processing

    Science.gov (United States)

    Fernández-González, Daniel; Martín-Duarte, Ramón; Ruiz-Bustinza, Íñigo; Mochón, Javier; González-Gasca, Carmen; Verdeja, Luis Felipe

    2016-06-01

    Blast furnace operators expect to get sinter with homogenous and regular properties (chemical and mechanical), necessary to ensure regular blast furnace operation. Blends for sintering also include several iron by-products and other wastes that are obtained in different processes inside the steelworks. Due to their source, the availability of such materials is not always consistent, but their total production should be consumed in the sintering process, to both save money and recycle wastes. The main scope of this paper is to obtain the least expensive iron ore blend for the sintering process, which will provide suitable chemical and mechanical features for the homogeneous and regular operation of the blast furnace. The systematic use of statistical tools was employed to analyze historical data, including linear and partial correlations applied to the data and fuzzy clustering based on the Sugeno Fuzzy Inference System to establish relationships among the available variables.

  5. Design, analysis, operation, and advanced control of hybrid renewable energy systems

    Science.gov (United States)

    Whiteman, Zachary S.

    Because using non-renewable energy systems (e.g., coal-powered co-generation power plants) to generate electricity is an unsustainable, environmentally hazardous practice, it is important to develop cost-effective and reliable renewable energy systems, such as photovoltaics (PVs), wind turbines (WTs), and fuel cells (FCs). Non-renewable energy systems, however, are currently less expensive than individual renewable energy systems (IRESs). Furthermore, IRESs based on intermittent natural resources (e.g., solar irradiance and wind) are incapable of meeting continuous energy demands. Such shortcomings can be mitigated by judiciously combining two or more complementary IRESs to form a hybrid renewable energy system (HRES). Although previous research efforts focused on the design, operation, and control of HRESs has proven useful, no prior HRES research endeavor has taken a systematic and comprehensive approach towards establishing guidelines by which HRESs should be designed, operated, and controlled. The overall goal of this dissertation, therefore, is to establish the principles governing the design, operation, and control of HRESs resulting in cost-effective and reliable energy solutions for stationary and mobile applications. To achieve this goal, we developed and demonstrated four separate HRES principles. Rational selection of HRES type: HRES components and their sizes should be rationally selected using knowledge of component costs, availability of renewable energy resources, and expected power demands of the application. HRES design: by default, the components of a HRES should be arranged in parallel for increased efficiency and reliability. However, a series HRES design may be preferred depending on the operational considerations of the HRES components. HRES control strategy selection: the choice of HRES control strategy depends on the dynamics of HRES components, their operational considerations, and the practical limitations of the HRES end-use. HRES data

  6. Multi-center evaluation of post-operative morbidity and mortality after optimal cytoreductive surgery for advanced ovarian cancer.

    Directory of Open Access Journals (Sweden)

    Arash Rafii

    Full Text Available PURPOSE: While optimal cytoreduction is the standard of care for advanced ovarian cancer, the related post-operative morbidity has not been clearly documented outside pioneering centers. Indeed most of the studies are monocentric with inclusions over several years inducing heterogeneity in techniques and goals of surgery. We assessed the morbidity of optimal cytoreduction surgery for advanced ovarian cancer within a short inclusion period in 6 referral centers dedicated to achieve complete cytoreduction. PATIENTS AND METHODS: The 30 last optimal debulking surgeries of 6 cancer centers were included. Inclusion criteria included: stage IIIc- IV ovarian cancer and optimal surgery performed at the site of inclusion. All post-operative complications within 30 days of surgery were recorded and graded using the Memorial secondary events grading system. Student-t, Chi2 and non-parametric statistical tests were performed. RESULTS: 180 patients were included. There was no demographic differences between the centers. 63 patients underwent surgery including intestinal resections (58 recto-sigmoid resection, 24 diaphragmatic resections, 17 splenectomies. 61 patients presented complications; One patient died post-operatively. Major (grade 3-5 complications requiring subsequent surgeries occurred in 21 patients (11.5%. 76% of patients with a major complication had undergone an ultraradical surgery (P = 0.004. CONCLUSION: While ultraradical surgery may result in complete resection of peritoneal disease in advanced ovarian cancer, the associated complication rate is not negligible. Patients should be carefully evaluated and the timing of their surgery optimized in order to avoid major complications.

  7. Control of a burning tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Burmeister, R.E.; Mandrekas, J.; Stacey, W.M.

    1993-03-01

    This report is a review of the literature relevant to the control of the thermonuclear burn in a tokamak plasma. Some basic tokamak phenomena are reviewed, and then control by modulation of auxiliary heating and fueling is discussed. Other possible control methods such as magnetic ripple, plasma compression, and impurity injection as well as more recent proposed methods such as divertor biasing and L- to H-mode transition are also reviewed. The applications of modern control theory to the tokamak burn control problem are presented. The control results are summarized and areas of further research are identified.

  8. Expansion of Michigan EOR Operations Using Advanced Amine Technology at a 600 MW Project Wolverine Carbon Capture and Storage Project

    Energy Technology Data Exchange (ETDEWEB)

    H Hoffman; Y kishinevsky; S. Wu; R. Pardini; E. Tripp; D. Barnes

    2010-06-16

    Wolverine Power Supply Cooperative Inc, a member owned cooperative utility based in Cadillac Michigan, proposes to demonstrate the capture, beneficial utilization and storage of CO{sub 2} in the expansion of existing Enhanced Oil Recovery operations. This project is being proposed in response to the US Department of Energy Solicitation DE-FOA-0000015 Section III D, 'Large Scale Industrial CCS projects from Industrial Sources' Technology Area 1. The project will remove 1,000 metric tons per day of CO{sub 2} from the Wolverine Clean Energy Venture 600 MW CFB power plant owned and operated by WPC. CO{sub 2} from the flue gas will be captured using Hitachi's CO{sub 2} capture system and advanced amine technology. The capture system with the advanced amine-based solvent supplied by Hitachi is expected to significantly reduce the cost and energy requirements of CO{sub 2} capture compared to current technologies. The captured CO{sub 2} will be compressed and transported for Enhanced Oil Recovery and CO{sub 2} storage purposes. Enhanced Oil Recovery is a proven concept, widely used to recover otherwise inaccessible petroleum reserves. While post-combustion CO{sub 2} capture technologies have been tested at the pilot scale on coal power plant flue gas, they have not yet been demonstrated at a commercial scale and integrated with EOR and storage operations. Amine-based CO{sub 2} capture is the leading technology expected to be available commercially within this decade to enable CCS for utility and industrial facilities firing coal and waste fuels such as petroleum coke. However, traditional CO{sub 2} capture process utilizing commercial amine solvents is very energy intensive for regeneration and is also susceptible to solvent degradation by oxygen as well as SOx and NO{sub 2} in the flue gas, resulting in large operating costs. The large volume of combustion flue gas with its low CO{sub 2} concentration requires large equipment sizes, which together with the

  9. Including collisions in gyrokinetic tokamak and stellarator simulations

    Energy Technology Data Exchange (ETDEWEB)

    Kauffmann, Karla

    2012-04-10

    Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal

  10. Evaluation of two preoparative chemotherapy regimens for complete operability of advanced gastric adenocarcinoma: a clinical trial

    Directory of Open Access Journals (Sweden)

    S. Sedighi

    2006-08-01

    Full Text Available Background: This prospective phase III study was designed to compare the activity of two combinations chemotherapy drugs in advanced gastric adenocarcinoma Methods: In a double blinded clinical trial, From Jan. 2002 to Jan. 2005, ninety patients with advanced gastric adenocarcinoma were randomly assigned to 1 Cisplatin and continuous infusion of 5FU and Epirubicin (ECF, and 2 Cisplatin and continuous infusion of 5FU with Docetaxel (TCF. Reduction in tumor mass, overall survival (OS, time to progression (TTP, and safety were measured outcome. Results: About 90% of patients had stage III or IV disease and the most common sites of tumor spread were peritoneal surfaces, liver and Paraaortic lymph nodes in either group. The objective clinical response rate (more than 50% decreases in tumor mass was 38% and 43% in ECF and TCF group respectively. Global quality of life increased (p=0 002 and symptoms of pain and insomnia decreased after chemotherapy. Patients in TCF had more grade one or two skin reactions, neuropathy and diarrhea. Fourteen patients underwent surgery. Complete microscopic (R0 resection had done in two of ECF and six of TCF tumors (p=0.015. Two cases in TCF group showed complete pathologic response. Median TTP was nine months and 10 months in ECF and TCF group respectively. Median OS was 12 months in both groups. Conclusion: Although there wasn’t statistically significant difference regarded to clinical response or survival between two groups, TCF showed more complete pathologic response.

  11. The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

    International Nuclear Information System (INIS)

    The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support

  12. The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    1993-12-01

    The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model`s on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy`s theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support.

  13. Intelligent Systems and Advanced User Interfaces for Design, Operation, and Maintenance of Command Management Systems

    Science.gov (United States)

    Mitchell, Christine M.

    1998-01-01

    Historically Command Management Systems (CMS) have been large, expensive, spacecraft-specific software systems that were costly to build, operate, and maintain. Current and emerging hardware, software, and user interface technologies may offer an opportunity to facilitate the initial formulation and design of a spacecraft-specific CMS as well as a to develop a more generic or a set of core components for CMS systems. Current MOC (mission operations center) hardware and software include Unix workstations, the C/C++ and Java programming languages, and X and Java window interfaces representations. This configuration provides the power and flexibility to support sophisticated systems and intelligent user interfaces that exploit state-of-the-art technologies in human-machine systems engineering, decision making, artificial intelligence, and software engineering. One of the goals of this research is to explore the extent to which technologies developed in the research laboratory can be productively applied in a complex system such as spacecraft command management. Initial examination of some of the issues in CMS design and operation suggests that application of technologies such as intelligent planning, case-based reasoning, design and analysis tools from a human-machine systems engineering point of view (e.g., operator and designer models) and human-computer interaction tools, (e.g., graphics, visualization, and animation), may provide significant savings in the design, operation, and maintenance of a spacecraft-specific CMS as well as continuity for CMS design and development across spacecraft with varying needs. The savings in this case is in software reuse at all stages of the software engineering process.

  14. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    D Raju; R Jha; P K Kaw; S K Mattoo; Y C Saxena; Aditya Team

    2000-11-01

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as the discharge progresses. It is observed that during the current rise phase, current perturbation undergoes transition from = 5 poloidal structure to = 4 and then to = 3. At the time of current termination, = 2 perturbation is observed. It is observed that the mode frequency remains nearly constant (≈10 kHz) when poloidal mode structure changes from = 4 to = 2. This may be either an indication of mode coupling or a consequences of changes in the plasma electron temperature and density scale length.

  15. Benefits of intra-operative systemic chemotherapy during curative surgery in patients with locally advanced gastric cancer

    Institute of Scientific and Technical Information of China (English)

    MENG Qing-bin; YU Jian-chun; MA Zhi-qiang; KANG Wei-ming; ZHOU Li; YE Xin

    2013-01-01

    Background There is little information on the impact of intra-operative systemic chemotherapy on gastric cancer.The aim of this study was to identify prognostic factors in patients with locally advanced gastric cancer and undergoing curative resection,with a focus on evaluating survival benefits and tolerance of intra-operative systemic chemotherapy.Methods We retrospectively analyzed clinicopathological data for 264 consecutive patients who underwent curative resection for gastric cancer at Peking Union Medical College Hospital from January 2002 to January 2007.Survival curves were plotted using the Kaplan-Meier method and compared using log-rank tests.Univariate and multivariate analyses were performed with the Cox proportional hazard model.Results Patients who received intra-operative systemic chemotherapy had higher 5-year overall survival and 5-year disease-free survival rates (P=0.019 and 0.010,respectively) than patients who did not receive intra-operative systemic chemotherapy.In the subgroup analysis,systemic intra-operative chemotherapy benefited the 5-year overall survival and disease-free survival rates for patients with cancer of stage pTNM ⅠB-ⅢB,but not stage pTNM ⅢC.Patients who received intra-operative systemic chemotherapy in combination with post-operative chemotherapy had higher 5-year overall survival and 5-year disease-free survival rates (P=0.046 and 0.021,respectively) than patients who only received postoperative chemotherapy.However,the difference in these rates between patients who received only intra-operative systemic chemotherapy and patients who only received curative surgery was not statistically significant (P=0.150 and 0.170,respectively).Multivariate analyses showed that intra-operative systemic chemotherapy was a favorable prognostic factor for the overall survival and disease-free survival rates (P =0.048 and 0.023,respectively).No grade 4 toxicities related to intra-operative systemic chemotherapy were recorded within the

  16. Advanced energy design and operation technologies research: Recommendations for a US Department of Energy multiyear program plan

    Energy Technology Data Exchange (ETDEWEB)

    Brambley, M.R.; Crawley, D.B.; Hostetler, D.D.; Stratton, R.C.; Addision, M.S.; Deringer, J.J.; Hall, J.D.; Selkowitz, S.E.

    1988-12-01

    This document describes recommendations for a multiyear plan developed for the US Department of Energy (DOE) as part of the Advanced Energy Design and Operation Technologies (AEDOT) project. The plan is an outgrowth of earlier planning activities conducted for DOE as part of design process research under the Building System Integration Program (BSIP). The proposed research will produce intelligent computer-based design and operation technologies for commercial buildings. In this document, the concept is explained, the need for these new computer-based environments is discussed, the benefits are described, and a plan for developing the AEDOT technologies is presented for the 9-year period beginning FY 1989. 45 refs., 37 figs., 9 tabs.

  17. First Results from the Lithium Tokamak eXperiment

    International Nuclear Information System (INIS)

    Full text: The Lithium Tokamak eXperiment (LTX) is a newly commissioned, modest-scale spherical tokamak with R =0.4 m, a =0.26 m, and elongation of 1.5. Design targets are a toroidal field of 3.2 kG, plasma current up to 400 kA, and a discharge duration of order 100 msec. LTX is the first tokamak designed to investigate modifications to equilibrium and transport when global recycling is reduced to 10 - 20 %. To reduce recycling, LTX is fitted with a 1 cm thick heated (300 deg. C) copper shell, conformal to the last closed flux surface, over 85% of the plasma surface area. The plasma-facing surface of the shell will be evaporatively coated with a thin (< 100 micron) layer of molten lithium, retained by surface tension. The shell is replaceable, and a second version has been constructed, which was plasma-sprayed with 100 - 200 microns of molybdenum to form a high-Z substrate for subsequent coating with lithium. After the installation of the second shell (in 2011), a high temperature (500 - 600 deg. C) operating phase for LTX is planned. LTX is the first tokamak designed to operate with a full hot high-Z wall, near the projected operating temperature for reactor PFCs. The engineering design and construction of the hot high-Z shell, as well as the vessel and diagnostics to tolerate both lithium and 500 deg. C internal components, will be discussed. LTX will employ short-pulse fueling with a new hydrogen molecular cluster injector, to transiently eliminate edge gas (between puffs). This fueling system will be briefly discussed. Diagnostics include single-pulse multipoint Thomson scattering, Lyman alpha arrays, microwave interferometers, spectrometers, and an edge Langmuir probe. LTX is now progressing through the shakedown phase, and first operation with a liquid lithium film wall is scheduled for Spring 2010. Later in 2010, a new diagnostic (Digital Holography) for core density variations will be tested on LTX. In 2011 a 5 A, 20 kV, 1 second hydrogen neutral beam

  18. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  19. Wide range operation of advanced low NOx aircraft gas turbine combustors

    Science.gov (United States)

    Roberts, P. B.; Fiorito, R. J.; Butze, H. F.

    1978-01-01

    The paper summarizes the results of an experimental test rig program designed to define and demonstrates techniques which would allow the jet-induced circulation and vortex air blast combustors to operate stably with acceptable emissions at simulated engine idle without compromise to the low NOx emissions under the high-altitude supersonic cruise condition. The discussion focuses on the test results of the key combustor modifications for both the simulated engine idle and cruise conditions. Several range-augmentation techniques are demonstrated that allow the lean-reaction premixed aircraft gas turbine combustor to operate with low NOx emissons at engine cruise and acceptable CO and UHC levels at engine idle. These techniques involve several combinations, including variable geometry and fuel switching designs.

  20. A Distributed Computing Architecture to Enable Advances in Field Operations and Management of Distributed Infrastructure

    OpenAIRE

    Khan, Kashif,

    2012-01-01

    Distributed infrastructures (e.g., water networks and electric Grids) are difficult to manage due to their scale, lack of accessibility, complexity, ageing and uncertainties in knowledge of their structure. In addition they are subject to loads that can be highly variable and unpredictable and to accidental events such as component failure, leakage and malicious tampering. To support in-field operations and central management of these infrastructures, the availability of consistent and up-to-...