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Sample records for advanced tokamak operation

  1. Advanced tokamak operating modes in TPX and ITER

    International Nuclear Information System (INIS)

    A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER

  2. 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

  3. Real-Time Profile Control for Advanced Tokamak Operation

    International Nuclear Information System (INIS)

    Simultaneous control of the plasma shape, the magnetic and kinetic plasma profiles (such as the safety factor, q(x), and gyro-normalized temperature gradient, ρTe*;(x), respectively) and the boundary flux is being investigated on JET, and has potential applications in the operation of ITER steady state advanced tokamak discharges. The control of radially distributed parameters was achieved for the first time on JET in 2004 [1-4]. The controller was based on the static plasma response only. The approach newly implemented on JET aims to use a dynamical plasma model, all the available heating and current drive (H and CD) systems, and the poloidal field (PF) system in an optimal way to achieve a set of requested magnetic and kinetic profiles. This paper describes the new model-based optimal profile controller which has been tested during the last 2007 experimental campaign. The controller aims to use the combination of heating and current drive systems - and optionally the PF system. First experimental results of current profile control obtained during the last 2007 JET campaign are presented

  4. 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.

  5. Development on JET of Advanced Tokamak Operations for ITER

    International Nuclear Information System (INIS)

    Recent research on Advanced Tokamak in JET has focused on scenarii with both monotonic and reversed shear q profiles having plasma parameters as relevant as possible for extrapolation to ITER. Wide ITBs, R∼3.7m, are formed at ITER relevant triangularity δ∼0.45, with ne/nG∼60% and ELMs moderated by Ne injection. At higher current (IP≤3.5MA, δ∼0.25) wide ITBs sitting at R≥ 3.5m (positive shear region) have been developed, generally MHD events terminate these barrier otherwise limited in strength by power availability. ITBs with core density close to Greenwald value are obtained with plasma target preformed by opportune timing of LHCD, pellet injection and small amount of NBI power. ITB start with toroidal rotation 4 times lower than the standard NBI heated ITBs. Full CD is achieved in reversed shear ITBs at 3T/1.8 MA, by using 10MW NBI, 5MW ICRH and 3MW LH. Wide ITBs located at R=3.6m, without impurity accumulation and type-III ELMs edge can be sustained for a time close to neo-classical resistive time. These discharges have been extended to the maximum duration allowed by subsystems (20s) with the JET record of injected energy: E∼330 MJ. Integrated control of pressure and current profile isit; feature used in these discharges. Central ICRF mode conversion electron heating, added to about 14MW NBI power, produced impressive ITBs with equivalent QDT ∼ 0.25. Conversely ion ITBs are obtained with low torque injection, by ICRH 3He minority heating of ions, on pure LHCD electron ITBs. Similarity experiments between JET and AUG have compared the dynamics of ITBs and have been the starting point of Hybrid Scenarios activity, then developed at ρ* as low as ρ*∼3*10-3. The development of hybrid regime with dominant electron heating has also started. Injection of trace of tritium and a mixture of Ar/Ne allowed studying fuel and impurities transport in many of the explored AT scenarios. (author)

  6. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

  7. 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

  8. 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

  9. 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.

  10. 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.

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

    International Nuclear Information System (INIS)

    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

  12. 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)

  13. 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.

  14. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    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

  15. 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.

  16. 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

  17. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

  18. 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.

  19. Real-time control of the q-profile in JET for steady state advanced tokamak operation

    International Nuclear Information System (INIS)

    In order to simultaneously control the current and pressure profiles in high performance tokamak plasmas with internal transport barriers (ITB), a multi-variable model-based technique has been proposed. New algorithms using a truncated singular value decomposition (TSVD) of a linearized model operator and retaining the distributed nature of the system have been implemented in the JET control system. Their simplest versions have been applied to the control of the current density profile in reversed shear plasmas using three heating and current drive actuators (neutral beam injection, ion cyclotron resonant frequency heating and lower hybrid current drive). Successful control of the safety factor profile has been achieved in the quasi-steady-state, on a timescale of the order of the current redistribution time. How the TSVD algorithm will be used in the forthcoming campaigns for the simultaneous control of the current profile and of the ITB temperature gradient is discussed in some detail, but this has not yet been attempted in the present pioneering experiments. (author)

  20. 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

  1. Magnetohydrodynamic instability, feedback stabilization, and disruption study for the Korea superconducting tokamak advanced research tokamak

    International Nuclear Information System (INIS)

    Passive and active feedback stabilization schemes being considered in Korea Superconducting Tokamak Advanced Research (KSTAR) device for the stabilization of the resistive magnetohydrodynamic modes such as the resistive wall and the neoclassical tearing are briefly introduced. A short summary is also presented on the tokamak simulation results of disruption dynamics and load in the KSTAR tokamak obtained using the tokamak simulation code (TSC)

  2. A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

    OpenAIRE

    Mitsuru Kikuchi

    2010-01-01

    Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR) best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fund...

  3. 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

  4. Advanced scenarios for ITER operation

    International Nuclear Information System (INIS)

    In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving conditions required for a reactor. An international experiment, ITER is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of α-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence in ITER achieving its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios on the other hand use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimisation of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the result obtained in experiments

  5. 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

  6. A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

    Directory of Open Access Journals (Sweden)

    Mitsuru Kikuchi

    2010-11-01

    Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

  7. KTM Tokamak operation scenarios software infrastructure

    International Nuclear Information System (INIS)

    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.

  8. 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.

  9. 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

  10. LIDAR Thomson scattering for advanced tokamaks. Final report

    International Nuclear Information System (INIS)

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured

  11. 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

  12. High performance operational limits of tokamak and helical systems

    International Nuclear Information System (INIS)

    The plasma operational boundaries of tokamak and helical systems are surveyed and compared with each other. Global confinement scaling laws are similar and gyro-Bohm like, however, local transport process is different due to sawtooth oscillations in tokamaks and ripple transport loss in helical systems. As for stability limits, achievable tokamak beta is explained by ideal or resistive MHD theories. On the other hand, beta values obtained so far in helical system are beyond ideal Mercier mode limits. Density limits in tokamak are often related to the coupling between radiation collapse and disruptive MHD instabilities, but the slow radiation collapse is dominant in the helical system. The pulse length of both tokamak and helical systems is on the order of hours in small machines, and the longer-pulsed good-confinement plasma operations compatible with radiative divertors are anticipated in both systems in the future. (author)

  13. 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

  14. 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.

  15. 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. PMID:26724028

  16. Continuous tokamak operation with an internal transformer

    International Nuclear Information System (INIS)

    A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia time scales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors

  17. Continuous tokamak operation with an internal transformer

    International Nuclear Information System (INIS)

    A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia timescales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors

  18. Operating tokamaks with steady-state toroidal current

    International Nuclear Information System (INIS)

    Continuous operation of a tokamak requires, among other things, a means of continuously providing the toroidal current. Various methods have been proposed to provide this current including methods which utilize radio-frequency waves in any of several frequency regimes. Here we elaborate on the prospects of incorporating these current-drive techniques in tokamak reactors, concentrating on the theoretical minimization of the power requirements

  19. Design of selected subsystems for COMPASS tokamak operation

    Czech Academy of Sciences Publication Activity Database

    Janky, Filip; Pereira, T.; Hron, Martin; Pánek, Radomír; Fernandes, H.

    Aix-en-Provence : IAEA, 2009. s. 80-80. ISBN N. [Seventh IAEATechnical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research. 15.06.2009-19.06.2009, Aix-en-Provence] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * Compass * machine control * tokamak operation Subject RIV: BL - Plasma and Gas Discharge Physics http://www-fusion-magnetique.cea.fr/tmiaea2009/ website /data/articles/000080.pdf

  20. The ARIES Advanced and Conservative Tokamak Power Plant Study

    International Nuclear Information System (INIS)

    Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5,a βtotalN of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m2. The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced-activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βtotalN of 2.5, an H98 of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here

  1. 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

  2. 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

  3. Lithium-cooled blankets for advanced tokamaks

    International Nuclear Information System (INIS)

    The main objective of the Tokamak Power System Studies (TPSS) at Argonne National Lab. during fiscal year 1985 was to explore innovative design concepts that have the potential for significant enhancement of the attractiveness of a tokamak-based power plant. Activities in the area of plasma engineering resulted in a reference reactor concept, which served as a model for the impurity control and first-wall/blanket/shield studies. The liquid-metal-cooled first-wall/blanket/shield design activity was centered around the vanadium alloy structure and liquid-lithium coolant leading blanket concept as identified by the Blanket Comparison and Selection Study (BCSS). A ferritic steel structure and a LiPb breeder were considered as backup options. The magnetohydrodynamics (MHD) effects associated with self-cooled liquid-metal blanket/first-wall systems are substantially reduced by the lower magnetic fields required for higher plasmas, the lower neutron wall loading resulting from reduced power output, and the smaller reactor size of the TPSS model reactor. Therefore, improved performance characteristics of self-cooled liquid-metal blanket concepts are achievable mainly because the design constraints are more relaxed compared to the BCSS guidelines. Key aspects of the designs evaluated in the current study include the following: (1) design simplicity; (2) use of the first wall as an impurity control device; (3) modular first-wall/blanket/reflector/shield construction; and (4) integrated first-wall/blanket/reflector/shield

  4. A concept for next step advanced tokamak fusion device

    International Nuclear Information System (INIS)

    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 = 1.2 m 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 insider the envisioned tokamak torus, with retrieved space in the direction of plasma minor radius, will be available for technological adjustments and maneuvering to facilitate implementation of engineering instrumentation and real

  5. 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

  6. Advanced fusion technologies developed for JT-60 superconducting tokamak

    International Nuclear Information System (INIS)

    The modification of JT-60U is planned as a full superconducting tokamak (JT-60SC). The objectives of the JT-60SC program are to establish scientific and technological bases for the steady-state operation of high performance plasmas and utilization of reduced-activation materials in economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to DEMO reactor have been developed in the superconducting magnet technology and plasma facing components for the design of JT-60SC. To achieve a high current density in a superconducting strand, Nb3Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFC) of JT-60SC. The R and D to demonstrate applicability of Nb3Al conductor to the TFC by a react-and-wind technique have been carried out using a full-size Nb3Al conductor. A full-size NbTi conductor with low AC loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the CFC target was successfully demonstrated on the electron beam irradiation stand. (author)

  7. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  8. Development of a free boundary Tokamak Equilibrium Solver (TES) for Advanced Study of Tokamak Equilibria

    CERN Document Server

    Jeon, Y M

    2015-01-01

    A free-boundary Tokamak Equilibrium Solver (TES), developed for advanced study of tokamak equilibra, is described with two distinctive features. One is a generalized method to resolve the intrinsic axisymmetric instability, which is encountered after all in equilibrium calculation with a free-boundary condition. The other is an extension to deal with a new divertor geometry such as snowflake or X divertors. For validations, the uniqueness of a solution is confirmed by the independence on variations of computational domain, the mathematical correctness and accuracy of equilibrium profiles are checked by a direct comparison with an analytic equilibrium known as a generalized Solovev equilibrium, and the governing force balance relation is tested by examining the intrinsic axisymmetric instabilities. As a valuable application, a snowflake equilibrium that requires a second order zero of the poloidal magnetic field is discussed in the circumstance of KSTAR coil system.

  9. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

  10. Towards steady-state tokamak operation with double transport barriers

    International Nuclear Information System (INIS)

    Internal Transport Barriers characteristic for the Optimised Shear regime and an edge transport barrier of an ELMy H-mode regime have been superposed in the Double Barrier mode. In DT discharges the Double Barrier mode has resulted in 50% higher fusion power output and a factor 2 higher fusion gain Q than in conventional sawtoothing steady-state ELMy H-mode plasmas. Steady-state conditions in temperature and density profiles have been approached in Double Barrier discharges in deuterium. The Double Barrier mode has been routinely established in the new Gas Box divertor configuration on JET. Off-axis LHCD has been used for current profile control during the high performance phase. In preparation of a new DTE2 campaign on JET the potential of the Double Barrier mode for sustained high fusion performance has been explored in modelling studies. Steady-state operation on ITER has been studied in transport code modelling for Advanced Tokamak scenarios in the Double Barrier mode. (author)

  11. Overview of the Microwave Tokamak Experiment operation and developments

    International Nuclear Information System (INIS)

    At Lawrence Livermore National Laboratory (LLNL), we assembled and presently operate the Microwave Tokamak Experiment (MTX) to demonstrate the feasibility of using intense microwave pulses (up to 6 GW peak power) from a free electron laser (FEL) to provide electron cyclotron heating (ECH) for use in tokamaks, particularly high field machines. The MTX consists primarily of the ALCATOR C tokamak and power supplies from MIT, along with FEL; the FEL is made up of the ETA-II linear induction accelerator and the IMP steady-state wiggler. A four-barrel pellet injector was added to the tokamak to produce peaked density profiles. The tokamak operations started in November 1988, with full duration plasmas being obtained at a toroidal field of both 5 and 9 tesla. Initial results were obtained with the single pulse 140 GHz FEL at peak power levels of 200 to 400 MW late in 1989. Due to excessive transverse electron beam motion, and arcing in the accelerator cells, the accelerator was modified. These modifications have been successfully tested on a small portion of the rebuilt accelerator and have been incorporated in the remaining portion of the accelerator. A 140 GHz, 400 kW gyrotron was used to perform preliminary heating experiments during the fall of 1990. This same gyrotron system is serving as the master oscillator for the burst mode FEL. The new IMP steady state wiggler will be used to produce the high power microwaves for the burst mode. The FEL construction has been completed, and it will be used for heating experiments scheduled for this fall. This paper describes the recent experimental operations. It also briefly outlines the additions and improvements to the experiment, which are described in more detail in companion papers at this conference

  12. 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)

  13. Halo current diagnostic system of experimental advanced superconducting tokamak.

    Science.gov (United States)

    Chen, D L; Shen, B; Granetz, R S; Sun, Y; Qian, J P; Wang, Y; Xiao, B J

    2015-10-01

    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. PMID:26520954

  14. Operating parameters of a reactor for early demonstration of electric power generation and the expansion by realization of advanced tokamak plasma

    International Nuclear Information System (INIS)

    Beam driven stable equilibria for a conceptual reactor, Demo-CREST, which was designed for early demonstration of electric power generation, has been investigated. Considering current profiles driven by neutral beams, the attainable normalized beta βN with a stabilization wall is about 3.4 with a normal shear (NS). With reversed shear (RS), a higher βN is attainable. The stable equilibria up to 4.0 can be sustained by a couple of On- and Off-axis beams. In the range of 1.9 N N = 1.9 which is the base design point of Demo-CREST. In the case of RS operation with βN 4.0, the density ratio to the Greenwald limit can be maintain at about unity if high temperature operation with Te > 20 kV is allowable. (author)

  15. Thermal Event Recognition Applied to Tokamak Protection during Plasma Operation

    OpenAIRE

    Martin, Vincent; Bremond, François; Travere, Jean-Marcel; Moncada, Victor; Dunand, Gwenaël

    2009-01-01

    Magnetic confinement fusion reactors are complex devices where a large amount of energy is required to make the fusion reactions happen. In such experimental conditions, the Plasma Facing Components (PFC) are subjected to high heat fluxes. In current tokamaks like Tore Supra, infrared thermographic diagnostics based on image analysis and feedback control are used to measure and monitor the heating of the PFC during plasma operation. The system consists in detecting high increase of the IR lum...

  16. 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...

  17. 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

  18. Operational limits and disruptions in tokamaks

    International Nuclear Information System (INIS)

    This report contains a concise description on status of knowledge in the area of operational limits and disruption characteristics, based on the results of workshops held during the ITER Definition Phase (1988). (author)

  19. The optimized li advanced tokamak scenario with high bootstrap current fraction

    International Nuclear Information System (INIS)

    Equilibrium and stability analyses have identified a class of tokamak configurations with conventional safety factor profiles (q0∼qmin approx-gt 1) at moderately high li(li∼1.0), and high normalized β(βN∼3.5 - 4.0), that are stable to the ideal n=1 kink without the requirement of wall stabilization. In contrast to previously identified high li, high βN equilibria, these configurations have high bootstrap current fractions (fBS∼50%- 70%); they require only modest central current drive for maintaining steady state and are therefore compatible with advanced tokamak (AT) operation. Strong plasma shaping is crucial for achieving the high β and high bootstrap fraction simultaneously. copyright 1999 American Institute of Physics

  20. 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.)

  1. 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...... repetition frequency increased immediately. The frequency and amplitude of type-III ELMs can be effectively influenced by puffing impurity argon gas....

  2. Development of operation scenarios with high bootstrap, negative shear configuration for large-aspect-ratio (LAR) bootstrap tokamak KT-2 at KAERI

    International Nuclear Information System (INIS)

    Through time dependent transport simulation, the authors have developed an operation scenario with high bootstrap, negative shear configuration for KT-2 tokamak, so that investigations of the so-called advanced tokamak operations are possible. In this study, they have concentrated on calculation of power requirements of heating and current drive system, MHD (Magnetohydrodynamic) stability and compatibility with PF (Poloidal Field) coil system. The study shows that high bootstrap (>70%), MHD stable (up to βN ≤ 3.9) operation scenario with negative shear configuration is possible within KT-2 tokamak design specification

  3. 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

  4. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak.

    Science.gov (United States)

    Cui, Z Q; Chen, Z J; Xie, X F; Peng, X Y; Hu, Z M; Du, T F; Ge, L J; Zhang, X; Yuan, X; Xia, Z W; Hu, L Q; Zhong, G Q; Lin, S Y; Wan, B N; Fan, T S; Chen, J X; Li, X Q; Zhang, G H

    2014-11-01

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G. PMID:25430242

  5. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G

  6. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    Science.gov (United States)

    Cui, Z. Q.; Chen, Z. J.; Xie, X. F.; Peng, X. Y.; Hu, Z. M.; Du, T. F.; Ge, L. J.; Zhang, X.; Yuan, X.; Xia, Z. W.; Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N.; Fan, T. S.; Chen, J. X.; Li, X. Q.; Zhang, G. H.

    2014-11-01

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G.

  7. 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)

  8. Collective effects and self-consistent energetic particle dynamics in advanced tokamaks

    International Nuclear Information System (INIS)

    In the present paper, we address the issue of fast ion and fusion product transport in conditions that are typically relevant for burning plasmas operating in so called Advanced Tokamak regimes. Our results have direct implications, e.g., on the choice of current profiles for ITER steady state operations. We demonstrate that in a Tokamak equilibrium with hollow-q profile, in general, two types of EPM (Energetic Particle Modes gap modes may exist near the minimum-q surface (q = q0), characterized by opposite signature in frequency: one with upwards chirping frequency and the other with downward chirping frequency as q 0 drops. It is shown that EPM gap modes are described by the same dispersion relation of the usual resonant EPMs and that they can indeed be considered as the same mode with, however, different dominant damping mechanisms. This work also presents a discussion of EPM non-linear dynamics with respect to energetic ion transport in tokamaks with hollow q-profiles. Numerical simulations based on a Hybrid MHD-Gyrokinetic Code (HMGC), demonstrate that, above the EPM excitation threshold, fast radial redistribution of energetic ions takes place on a time scale that is proportional to the inverse EPM growth rate (typically ∼ 100τ R0/vA being the Alfven time). The rapid evolution of EPM mode structures and the associated fast ion transport is interpreted within the framework of the relay runner model for non-linear EPM dynamics. It is found that a sensitive parameter for tokamak equilibria with hollow-q profiles is q at the minimum-q surface, higher q corresponding to larger particle transport. This fact has clear implications on the choice of current profiles in a burning plasma. (author)

  9. 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

  10. A new remote control room for tokamak operations

    International Nuclear Information System (INIS)

    This paper presents a summary of a new remote tokamak control room constructed near the offices of DIII-D's scientific staff. This integrated system combines hardware, software, data, and control of the room (R-232) into a unified package that has been designed and constructed in a generic fashion so that it can be used with any tokamak operating worldwide. The room is approximately 300 ft2 and can accommodate up to 12 seated participants. Mounted on the wall facing each scientist are five 52″ LCD televisions and mounted to the wall on their right are six 24″ LCD monitors. Each seat has associated with it a 24″ monitor, network connection, and power and the scientist is either provided with a computer or they can use their own. The room has been used for operation of DIII-D, EAST, and KSTAR. Due to the long distances, data from EAST and KSTAR was brought back to local DIII-D computers in one large parallel network transfer and subsequently served to scientists in the remote control room to other US collaborators. This parallel data transfer allowed the data to be available to US participants between pulses making remote experimental participation highly effective.

  11. 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

  12. 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.

  13. Influence of helium puff on divertor asymmetry in experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, G. S.; Wang, L.; Wang, H. Q.; Ding, R.; Duan, Y. M.; Gan, K. F.; Shao, L. M.; Chen, L.; Yan, Ning; Zhang, W.; Chen, R.; Xiong, H.; Ding, S.; Hu, G. H.; Liu, Y. L.; Zhao, N.; Li, Y. L.; Gao, X.

    2014-01-01

    Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected into...

  14. ARIES-AT: An advanced tokamak, advanced technology fusion power plant

    International Nuclear Information System (INIS)

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant. Several avenues were pursued in order to arrive at plasmas with a higher β and better bootstrap alignment compared to ARIES-RS that led to plasmas with higher βN and β. Advanced technologies that are examined in detail include: (1) Possible improvements to the overall system by using high-temperature superconductors, (2) Innovative SiC blankets that lead to a high thermal cycle efficiency of ∼60%; and (3) Advanced manufacturing techniques which aim at producing near-finished products directly from raw material, resulting in low-cost, and reliable components. The 1000-MWe ARIES-AT design has a major radius of 5.4 m, minor radius of 1.3 M, a toroidal β of 9.2% (βN=6.0) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current drive power is 24 MW. The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (5c/kWh), which is competitive with those projected for other sources of energy. (author)

  15. 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 the value specified for Q = 10 in ITER reference scenario, under stationary conditions. Discharges have also been demonstrated in DIII-D with enhanced performance under stationary conditions that project to Q ∼ 10 for longer than 1 h in ITER at reduced current, if such a mode of operation can be realized in ITER. 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, edge localized mode 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 behaviour in the electron heat transport, rotation in plasmas in the absence of external torque, measurements in the edge pedestal region and plasma fuelling. Understanding these phenomena at a fundamental level contributes to development and ultimately the optimization of tokamak scenarios

  16. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

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

    Science.gov (United States)

    Qu, Hao; Zhang, Tao; Han, Xiang; Wen, Fei; Zhang, Shoubiao; Kong, Defeng; Wang, Yumin; Gao, Yu; Huang, Canbin; Cai, Jianqing; Gao, Xiang

    2015-08-01

    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. PMID:26329188

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

    International Nuclear Information System (INIS)

    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

  19. Advances in comprehensive gyrokinetic simulations of transport in tokamaks

    International Nuclear Information System (INIS)

    A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate core turbulent transport in actual experimental profiles and enable direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite β, equilibrium ExB shear stabilization, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius (ρ*) so as to treat the profile shear stabilization and nonlocal effects which can break gyroBohm scaling. The code operates in either a cyclic flux-tube limit (which allows only gyroBohm scaling) or globally with physical profile variation. Bohm scaling of DIII-D L-mode has been simulated with power flows matching experiment within error bars on the ion temperature gradient. Mechanisms for broken gyroBohm scaling, neoclassical ion flows embedded in turbulence, turbulent dynamos and profile corrugations, are illustrated. (author)

  20. Development of an alternating integrator for magnetic measurements for experimental advanced superconducting tokamak.

    Science.gov (United States)

    Liu, D M; Wan, B N; Zhao, W Z; Shen, B; He, Y G; Chen, B; Huang, J; Liu, H Q

    2014-11-01

    A high-performance integrator is one of the key electronic devices for reliably controlling plasma in the experimental advanced superconducting tokamak for long pulse operation. We once designed an integrator system of real-time drift compensation, which has a low integration drift. However, it is not feasible for really continuous operations due to capacitive leakage error and nonlinearity error. To solve the above-mentioned problems, this paper presents a new alternating integrator. In the new integrator, the integrator system of real-time drift compensation is adopted as one integral cell while two such integral cells work alternately. To achieve the alternate function, a Field Programmable Gate Array built in the digitizer is utilized. The performance test shows that the developed integrator with the integration time constant of 20 ms has a low integration drift (<15 mV) for 1000 s. PMID:25430391

  1. Density limits investigation and high density operation in EAST tokamak

    Science.gov (United States)

    Zheng, Xingwei; Li, Jiangang; Hu, Jiansheng; Liu, Haiqing; Jie, Yinxian; Wang, Shouxin; Li, Jiahong; Duan, Yanming; Li, Miaohui; Li, Yongchun; Zhang, Ling; Ye, Yang; Yang, Qingquan; Zhang, Tao; Cheng, Yingjie; Xu, Jichan; Wang, Liang; Xu, Liqing; Zhao, Hailin; Wang, Fudi; Lin, Shiyao; Wu, Bin; Lyu, Bo; Xu, Guosheng; Gao, Xiang; Shi, Tonghui; He, Kaiyang; Lan, Heng; Chu, Nan; Cao, Bin; Sun, Zhen; Zuo, Guizhong; Ren, Jun; Zhuang, Huidong; Li, Changzheng; Yuan, Xiaolin; Yu, Yaowei; Wang, Houyin; Chen, Yue; Wu, Jinhua; EAST Team

    2016-05-01

    Increasing the density in a tokamak is limited by the so-called density limit, which is generally performed as an appearance of disruption causing loss of plasma confinement, or a degradation of high confinement mode which could further lead to a H  →  L transition. The L-mode and H-mode density limit has been investigated in EAST tokamak. Experimental results suggest that density limits could be triggered by either edge cooling or excessive central radiation. The L-mode density limit disruption is generally triggered by edge cooling, which leads to the current profile shrinkage and then destabilizes a 2/1 tearing mode, ultimately resulting in a disruption. The L-mode density limit scaling agrees well with the Greenwald limit in EAST. The observed H-mode density limit in EAST is an operational-space limit with a value of 0.8∼ 0.9{{n}\\text{GW}} . High density H-mode heated by neutral beam injection (NBI) and lower hybrid current drive (LHCD) are analyzed, respectively. The constancy of the edge density gradients in H-mode indicates a critical limit caused perhaps by e.g. ballooning induced transport. The maximum density is accessed at the H  →  L transition which is generally caused by the excessive core radiation due to high Z impurities (Fe, Cu). Operating at a high density (>2.8× {{10}19} {{\\text{m}}-3} ) is favorable for suppressing the beam shine through NBI. High density H-mode up to 5.3× {{10}19}{{\\text{m}}-3}~≤ft(∼ 0.8{{n}\\text{GW}}\\right) could be sustained by 2 MW 4.6 GHz LHCD alone, and its current drive efficiency is studied. Statistics show that good control of impurities and recycling facilitate high density operation. With careful control of these factors, high density up to 0.93{{n}\\text{GW}} stable H-mode operation was carried out heated by 1.7 MW LHCD and 1.9 MW ion cyclotron resonance heating with supersonic molecular beam injection fueling.

  2. High density operation on the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    The structure of the operation region has been studied in the HT-7 superconducting tokamak, and progress on the extension of the HT-7 ohmic discharge operation region is reported. A density corresponding to 1.2 times the Greenwald limit was achieved by RF boronization. The density limit appears to be connected to the impurity content and the edge parameters, so the best results are obtained with very clean plasmas and peaked electron density profiles. The peaking factors of electron density profiles for different current and line averaged densities were observed. The density behaviour and the fuelling efficiency for gas puffing (20-30%), pellet injection (70-80%) and molecular beam injection (40-50%) were studied. The core crash sawteeth and MHD behaviour, which were induced by an injected pellet, were observed and the events correlated with the change of current profile and reversed magnetic shear. The MARFE phenomena on HT-7 are summarized. The best correlation has been found between the total input ohmic power and the product of the edge line averaged density and Zeff. HT-7 could be easily operated in the high density region MARFE-free using RF boronization. (author)

  3. 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.

  4. Advanced tokamak physics experiments on DIII-D

    International Nuclear Information System (INIS)

    Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy confinement time (τE) and the plasma pressure (or beta βT = 2 μ0 /BT2) 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 βN Hiop ∼ 10 for 5 τE (∼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. 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 %.

  6. 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.

  7. Tokamak operation with high-Z plasma facing components

    International Nuclear Information System (INIS)

    Due to wall lifetime requirements and the problem of tritium co-deposition in hydrocarbon layers, a future burning plasma will most probably have a full high-Z wall. The prime candidate material is tungsten, which exhibits good thermo-mechanical properties and has a high energy threshold for physical sputtering. To investigate the reactor-compatibility of this wall material, ASDEX Upgrade is being converted into a full-tungsten coated tokamak in a step-by-step approach, with presently almost 70 % of W wall coverage. The effect of the reduction of primary carbon coverage on the plasma is so far moderate. Under tokamak conditions, carbon behaves like a recycling impurity, due to the deposition and re-erosion of soft hydrocarbon layers on the tungsten surface. During high density H-mode operation, the central tungsten concentrations remain typically low, i.e. well under 10-5. The situation is more critical in the improved H-mode or hybrid scenario. Here, the combination of hot edge conditions and peaked central density profiles result in high central tungsten concentrations of up to 10-4, which would be critical in a reactor. However, core electron density peaking is reduced by use of central ICR or ECR heating and thus in turn suppresses central tungsten accumulation. For extrapolation to reactor conditions, we need to separate the effects of the tungsten wall source, the penetration over the edge transport barrier (ETB) and the core transport with its strong neoclassical contribution. These issues are addressed by inspecting the tungsten behaviour in various discharge scenarios and parameters in ASDEX Upgrade. These include radiative cooling by medium-Z seed impurities and ELM frequency control by pellet injection to simulate a reactor plasma with small edge and divertor impurity radiation levels and a separatrix power flux close to the H-L threshold. Fast ions produced by NBI and ICR heating at the low field side appear to be an important tungsten erosion mechanism

  8. Advanced Tele-operation

    International Nuclear Information System (INIS)

    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

  9. 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

  10. 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)

  11. Advanced Neutron Source operating philosophy

    International Nuclear Information System (INIS)

    An operating philosophy and operations cost estimate were prepared to support the Conceptual Design Report for the Advanced Neutron Source (ANS), a new research reactor planned for the Oak Ridge National Laboratory (ORNL). The operating philosophy was part of the initial effort of the ANS Human Factors Program, was integrated into the conceptual design, and addressed operational issues such as remote vs local operation; control room layout and responsibility issues; role of the operator; simulation and training; staffing levels; and plant computer systems. This paper will report on the overall plans and purpose for the operations work, the results of the work done for conceptual design, and plans for future effort

  12. The impact of ripple on the design of advanced fuel tokamaks

    International Nuclear Information System (INIS)

    Enhanced ion transport due to toroidal field ripple is a concern in the design of tokamak power reactors. This concern is quantified for advanced fuel cycle applications where the simultaneous requirement of high /eta//tau/ and high T makes the constraint on deviation from axisymmetry especially severe

  13. TPX diagnostics for tokamak operation, plasma control and machine protection

    International Nuclear Information System (INIS)

    The diagnostics for TPX are at an early design phase, with emphasis on the diagnostic access interface with the major tokamak components. Account has to be taken of the very severe environment for diagnostic components located inside the vacuum vessel. The placement of subcontracts for the design and fabrication of the diagnostic systems is in process

  14. The ARIES-AT advanced tokamak, Advanced technology fusion power plant

    International Nuclear Information System (INIS)

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R and D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ x = 2.2) which is the result of a 'thinner' blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb-17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 deg. C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb-17Li to about 1000 deg. C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES

  15. 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

  16. Numerical study of Alfvén eigenmodes in the Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Alfvén eigenmodes in up-down asymmetric tokamak equilibria are studied by a new magnetohydrodynamic eigenvalue code. The code is verified with the NOVA code for the Solovév equilibrium and then is used to study Alfvén eigenmodes in a up-down asymmetric equilibrium of the Experimental Advanced Superconducting Tokamak. The frequency and mode structure of toroidicity-induced Alfvén eigenmodes are calculated. It is demonstrated numerically that up-down asymmetry induces phase variation in the eigenfunction across the major radius on the midplane

  17. 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.; Xu, G.S.; Guo, H. Y.; Cao, B.; Wang, H. Q.; Wang, L.; Yan, Ning; Xia, S. B.; Zhang, W.; Chen, R.; Chen, L.; Ding, S. Y.; Xiong, H.; Zhao, Y.; Wan, B. N.; Gong, X. Z.; Gao, X.

    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...... advanced superconducting tokamak (EAST). The two views are up-down symmetric about the midplane and separated by a toroidal angle of 66.6 degrees. A linear manifold with 16 holes apart by 10 mm is used to form helium gas cloud at the 130x130 mm (radial versus poloidal) objective plane. A fast camera is...

  18. Study of an advanced D-T tokamak fusion reactor with compact fusion advanced rankine (CFAR) cycle

    International Nuclear Information System (INIS)

    Recent progress of the CFAR (Compact Fusion Advanced Rankine) cycle concept for a D-T tokamak reactor is presented with emphasis on how an enthalpy extraction can be achieved by a nonequilibrium disk-type MHD generator. For the gas stagnation temperatures of 3,000 K, enthalpy extraction in excess of 50% is found to be achievable, leading to a 40% overall plant efficiency with application of recuperative heat cycle and recently advanced thermoelectric converters. About 6 ton/sec mercury flow is required to remove fusion energy while achieving the 3,000 K gas stagnation temperature prior to the MHD generator. Studies of plasma parameters in the steady-state operation regime subject to plasma physics constraints, the minimum power in the start up phase required for ignition, effects of MHD magnet to the plasma confining magnetic fields, neutron and microwave superheat, and mercury corrosion test of ceramic rods for 2,000 hours are also described. 14 refs., 6 figs., 1 tab

  19. ELM induced tungsten melting and its impact on tokamak operation

    Czech Academy of Sciences Publication Activity Database

    Coenen, J.W.; Arnoux, G.; Bazylev, B.; Matthews, G. F.; Jachmich, S.; Balboa, I.; Clever, M.; Dejarnac, Renaud; Coffey, I.; Corre, Y.; Devaux, S.; Frassinetti, L.; Gauthier, E.; Horáček, Jan; Knaup, M.; Komm, Michael; Krieger, K.; Marsen, S.; Meigs, A.; Mertens, Ph.; Pitts, R.A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.

    2015-01-01

    Roč. 463, August (2015), s. 78-84. ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] Institutional support: RVO:61389021 Keywords : plasma * tokamak Subject RIV: BG - Nuclear , Atomic and Molecular Physics, Colliders Impact factor: 1.865, year: 2014 http://www.sciencedirect.com/science/article/pii/S0022311514005960#

  20. Optimizing the beta limit in DIII-D advanced tokamak discharges

    International Nuclear Information System (INIS)

    Results are presented from comparisons of modeling and experiment in studies to assess the best choice of safety factor (q) profile, pressure profile and discharge shape for high beta, steady-state, noninductive advanced tokamak operation. This is motivated by the need for high qminβN to maximize the self-driven bootstrap current while maintaining high toroidal beta to increase fusion gain. Experiment and theory both show that increases in the achievable normalized beta (βN) can be obtained through broadening of the pressure profile and use of a symmetric double-null divertor shape. The general trend is for βN to decrease as the minimum q value (qmin) increases, but with a broadened pressure profile, βN = 4 is obtained with qmin ∼ 2 and qminβN increases with qmin. Modeling of equilibria with near 100% bootstrap current indicates that operation with βN ∼ 5 should be possible with a sufficiently broad pressure profile. (author)

  1. 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)

  2. Control and operation of the gas injection systems for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ok, E-mail: yokim@nfri.re.kr [National Fusion Research Institute, 113 Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Song, Jae In; Kim, Kwang Pyo; Chu, Yong; Park, Kap Rai; Kim, Hong Tack; Kim, Hak Kun; Lee, Kun Su [National Fusion Research Institute, 113 Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Kim, Yang Mo [Chungnam National Univ., Gung-dong, Yuseong-gu, Daejeon 305-764 (Korea, Republic of)

    2013-10-15

    Highlights: ► Three different types of gas injection systems were manufactured for KSTAR. ► They were prepared and tested for the KSTAR plasma operation. ► The test results show that they satisfied the design requirement. ► The calibration for all gas injection systems was successfully completed. -- Abstract: Several gas injection systems such as fueling, massive gas injection (MGI), supersonic molecular beam injection (SMBI) have been installed for the Korea Superconducting Tokamak Advanced Research (KSTAR). Piezoelectric valves with a built-in piezoelectric plate of disc-bender type were used for the precise feedback control of gas flow. The relationship between the displacement and blocking force of the piezoelectric plates is taken into account in designing and fabricating piezoelectric valves. The MGI system was installed for the mitigation of plasma disruption. The MGI valve is composed of two-volume chamber. The valve is equipped with an aluminum piston and an actuator coil of inductance 0.1 mH. The power supply unit using a capacitor discharge method has been developed in order to drive the MGI valve. The SMBI system consisting of LN{sub 2} cooling line, pulse valve, and its controller was operated for the mitigation of edge localized mode (ELM). The LN{sub 2} cooling line is used to increase the density of the injected gas. The status and specification of each gas injection system is presented and the operational characteristics are analyzed.

  3. Control and operation of the gas injection systems for KSTAR tokamak

    International Nuclear Information System (INIS)

    Highlights: ► Three different types of gas injection systems were manufactured for KSTAR. ► They were prepared and tested for the KSTAR plasma operation. ► The test results show that they satisfied the design requirement. ► The calibration for all gas injection systems was successfully completed. -- Abstract: Several gas injection systems such as fueling, massive gas injection (MGI), supersonic molecular beam injection (SMBI) have been installed for the Korea Superconducting Tokamak Advanced Research (KSTAR). Piezoelectric valves with a built-in piezoelectric plate of disc-bender type were used for the precise feedback control of gas flow. The relationship between the displacement and blocking force of the piezoelectric plates is taken into account in designing and fabricating piezoelectric valves. The MGI system was installed for the mitigation of plasma disruption. The MGI valve is composed of two-volume chamber. The valve is equipped with an aluminum piston and an actuator coil of inductance 0.1 mH. The power supply unit using a capacitor discharge method has been developed in order to drive the MGI valve. The SMBI system consisting of LN2 cooling line, pulse valve, and its controller was operated for the mitigation of edge localized mode (ELM). The LN2 cooling line is used to increase the density of the injected gas. The status and specification of each gas injection system is presented and the operational characteristics are analyzed

  4. Edge Thomson scattering diagnostic on COMPASS tokamak: installation, calibration, operation, improvements.

    Science.gov (United States)

    Bohm, P; Aftanas, M; Bilkova, P; Stefanikova, E; Mikulin, O; Melich, R; Janky, F; Havlicek, J; Sestak, D; Weinzettl, V; Stockel, J; Hron, M; Panek, R; Scannell, R; Frassinetti, L; Fassina, A; Naylor, G; Walsh, M J

    2014-11-01

    The core Thomson scattering diagnostic (TS) on the COMPASS tokamak was put in operation and reported earlier. Implementation of edge TS, with spatial resolution along the laser beam up to ∼1/100 of the tokamak minor radius, is presented now. The procedure for spatial calibration and alignment of both core and edge systems is described. Several further upgrades of the TS system, like a triggering unit and piezo motor driven vacuum window shutter, are introduced as well. The edge TS system, together with the core TS, is now in routine operation and provides electron temperature and density profiles. PMID:25430338

  5. 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

  6. High performance advanced tokamak regimes in DIII-D for next-step experiments

    International Nuclear Information System (INIS)

    Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results

  7. 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.)

  8. New steady-state quiescent high-confinement plasma in an experimental advanced superconducting tokamak.

    Science.gov (United States)

    Hu, J S; Sun, Z; Guo, H Y; Li, J G; Wan, B N; Wang, H Q; Ding, S Y; Xu, G S; Liang, Y F; Mansfield, D K; Maingi, R; Zou, X L; Wang, L; Ren, J; Zuo, G Z; Zhang, L; Duan, Y M; Shi, T H; Hu, L Q

    2015-02-01

    A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development. PMID:25699449

  9. 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

  10. Systems code studies for assessment and selection of operational points for a tokamak DEMO reactor

    International Nuclear Information System (INIS)

    A big step towards commercial fusion will be the commissioning of the first DEMO power plant. For tokamaks a steady state can only be achieved in advanced scenarios with a high bootstrap fraction and non-inductive current drive. For designing a reactor the parameter space is limited by physics constraints, which can hardly be computed alltogether on a sophisticated level. Thus empirical findings are extrapolated to larger machines by scaling laws in order to cast complex physics into computable macroscopic dependencies. Systems codes are used to predict stable operational points for given reactor scenarios and to perform parameter scans for optimized selection. Scenario specifications like divertor detachment and their effects on material lifetime and plasma stability need to be more properly represented in systems codes in order to define new criteria for the fine-selection of stable operational points, which can then be further investigated by highly sophisticated, but time-consuming codes. Therefore the scalings and equations in the code were cross-checked with latest experimental findings and a new module for lifetime estimations for the first Wall is being developed.

  11. Overview of JT-60U progress towards steady-state advanced tokamak

    International Nuclear Information System (INIS)

    Recent experimental results on steady state advanced tokamak (AT) research on JT-60U are presented with emphasis on longer time scale in comparison with characteristics time scales in plasmas. Towards this, modification on control in operation, heating and diagnostics systems have been done. As the results, ∼ 60 s Ip flat top and an ∼ 30 s H-mode are obtained. The long pulse modification has opened a door into a new domain for JT-60U. The high normalized beta (βN) of 2.3 is maintained for 22.3 s and 2.5 for 16.5 s in a high βp H-mode plasma. A standard ELMy H-mode plasma is also extended and change in wall recycling in such a longer time scale has been unveiled. Development and investigation of plasmas relevant to AT operation has been continued in former 15 s discharges as well in which higherNB power (≤ 10 s) is available. Higher βN ∼ 3 is maintained for 6.2 s in high βp H-mode plasmas. High bootstrap current fraction (fBS) of ∼ 75% is sustained for 7.4 s in an RS plasma. On NTM suppression by localized ECCD, ECRF injection preceding the mode saturation is found to be more effective to suppress the mode with less power compared to the injection after the mode saturated. The domain of the NTM suppression experiments is extended to the high βN regime, and effectiveness of m/n=3/2 mode suppression by ECCD is demonstrated at βN ∼ 2.5-3. Genuine center-solenoid less tokamak plasma start up is demonstrated. In a current hole region, it is shown that no scheme drives a current in any direction. Detailed measurement in both spatial and energy spaces of energetic ions showed dynamic change in the energetic ion profile at collective instabilities. Impact of toroidal plasma rotation on ELM behaviors is clarified in grassy ELM and QH domains. (author)

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

    Science.gov (United States)

    Liu, X; Zhao, H L; Liu, Y; Li, E Z; Han, X; Domier, C W; Luhmann, N C; Ti, A; Hu, L Q; Zhang, X D

    2014-09-01

    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. PMID:25273727

  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. Development of high thermal flux components for continuous operation in Tokamaks

    International Nuclear Information System (INIS)

    High heat flux plasma facing components are under development and appropriate experimental evaluations have been carried out in order to operate during cycles of several hundred seconds. In Tore Supra, a large tokamak with a plasma nominal duration in excess of 30 seconds, solutions are tested that could be later applied to the NET/ITER tokamak, where peaked heat flux values of 15 MW/m2 on the divertor plates are foreseen. The proposed concept is a swirl square tube design protected with brazed CFC flat tiles. Development programs and validation tests are presented. The tests results are compared with calculations

  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. 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.

  18. Understanding and Control of Transport in Advanced Tokamak Regimes in DIII-D

    International Nuclear Information System (INIS)

    Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection takes place when the heating power exceeds a threshold value of about 9 MW, contrasting to CO-NBI injection, where Pthreshold NH89 = 9 for 16 confinement times has been accomplished in a discharge combining an ELMing H-mode edge and an ITB, and exhibiting ion thermal transport down to 2-3 times neoclassical. The microinstabilities usually associated with ion thermal transport are predicted stable, implying that another mechanism limits performance. High frequency MHD activity is identified as the probable cause

  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. 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.

  2. 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.)

  3. 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

  4. 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)

  5. 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

  6. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    Energy Technology Data Exchange (ETDEWEB)

    Gentile, C.A.; LaMarche, P.H. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Anderson, J.L. [Los Alamos National Lab., NM (United States)

    1995-07-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During `` time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems.

  7. 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.

  8. 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. PMID:25725839

  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. 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

  11. An advanced plasma control system for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J. [General Atomics, San Diego, CA (United States); Lazarus, E. [Oak Ridge National Lab., TN (United States)

    1991-11-01

    An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as {beta}{sub p}, {ell}{sub i} and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 {mu}s intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 {mu}s.

  12. An Advanced Tokamak Fusion Nuclear Science Facility (FNSF-AT)

    Science.gov (United States)

    Chan, V. S.; Garofalo, A. M.; Stambaugh, R. D.

    2010-11-01

    A Fusion Development Facility (FDF) is a candidate for FNSF-AT. It is a compact steady-state machine of moderate gain that uses AT physics to provide the neutron fluence required for fusion nuclear science development. FDF is conceived as a double-null plasma with high elongation and triangularity, predicted to allow good confinement of high plasma pressure. Steady-state is achieved with high bootstrap current and radio frequency current drive. Neutral beam injection and 3D non-resonant magnetic field can provide edge plasma rotation for stabilization of MHD and access to Quiescent H-mode. The estimated power exhaust is somewhat lower than that of ITER because of higher core radiation and stronger tilting of the divertor plates. FDF is capable of further developing all elements of AT physics, qualifying them for an advanced performance DEMO. The latest concept has accounted for realistic neutron shielding and divertor implementation. Self-consistent evolution of the transport profiles and equilibrium will quantify the stability and confinement required to meet the FNS mission.

  13. 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

  14. 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

  15. 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.

  16. A new low drift integrator system for the Experiment Advanced Superconductor Tokamak

    International Nuclear Information System (INIS)

    A new type of the integrator system with the low drift characteristic has been developed to accommodate the long pulse plasma discharges on Experiment Advanced Superconductor Tokamak (EAST). The integrator system is composed of the Ethernet control module and the integral module which includes one integrator circuit, followed by two isolation circuits and two program-controlled amplifier circuits. It compensates automatically integration drift and is applied in real-time control. The performance test and the experimental results in plasma discharges show that the developed integrator system can meet the requirements of plasma control on the accuracy and noise level of the integrator in long pulse discharges.

  17. 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

  18. 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

  19. 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

  20. Containment flooding: Operating and advanced BWRs

    International Nuclear Information System (INIS)

    Currently, the emergency procedures, or even the strategies for operating advanced BWRs, have not been drafted and factored into the design of the advanced plants. This paper examines one particular strategy, containment flooding, that may be implemented in advanced BWRs. This paper then provides some of the lessons learned from currently operating plants regarding the effectiveness of this strategy

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, J.L. [Princeton Plasma Physics Laboratory (United States); Gentile, C. [Princeton Plasma Physics Laboratory (United States); Kalish, M. [Princeton Plasma Physics Laboratory (United States); Kamperschroer, J. [Princeton Plasma Physics Laboratory (United States); Kozub, T. [Princeton Plasma Physics Laboratory (United States); LaMarche, P. [Princeton Plasma Physics Laboratory (United States); Murray, H. [Princeton Plasma Physics Laboratory (United States); Nagy, A. [Princeton Plasma Physics Laboratory (United States); Raftopoulos, S. [Princeton Plasma Physics Laboratory (United States); Rossmassler, R. [Princeton Plasma Physics Laboratory (United States); Sissingh, R. [Princeton Plasma Physics Laboratory (United States); Swanson, J. [Princeton Plasma Physics Laboratory (United States); Tulipano, F. [Princeton Plasma Physics Laboratory (United States); Viola, M. [Princeton Plasma Physics Laboratory (United States); Voorhees, D. [Princeton Plasma Physics Laboratory (United States); Walters, R.T. [Princeton Plasma Physics Laboratory (United States)

    1995-03-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.).

  6. 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.

  7. 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

  8. 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.

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

    Science.gov (United States)

    Lampert, M; Anda, G; Czopf, A; Erdei, G; Guszejnov, D; Kovácsik, Á; Pokol, G I; Réfy, D; Nam, Y U; Zoletnik, S

    2015-07-01

    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. PMID:26233377

  10. The study of heat flux for disruption on experimental advanced superconducting tokamak

    Science.gov (United States)

    Yang, Zhendong; Fang, Jianan; Gong, Xianzu; Gan, Kaifu; Luo, Jiarong; Zhao, Hailin; Cui, Zhixue; Zhang, Bin; Chen, Meiwen

    2016-05-01

    Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dRsep = -2 cm, while it changes to upper single null (dRsep = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m2.

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

    International Nuclear Information System (INIS)

    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

  12. 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.

  13. A unified theory of tokamak transport via the generalized Balescu--Lenard collision operator

    International Nuclear Information System (INIS)

    A unified basis from which to study the transport of tokamaks at low collisionality is provided by specializing the ''generalized Balescu--Lenard'' collision operator to toridal geometry. Explicitly evaluating this operator, ripple, turbulent, and neoclassical transport coefficients are obtained, simply by further specializing the single operator to different particular classes of fluctuation wavelength and mode structure. For each class of fluctuations, the operator possesses a diffusive, test-particle contribution D, and in addition a dynamic drag term F, which makes the operator self-consistent, and whose presence is accordingly essential for the resultant fluxes to possess the appropriate conservation laws and symmetrics. These properties, well-known for axisymmetric transport, are demonstrated for one type of turbulent transport, chosen for definiteness, by explicit evaluation of both ''anomalous diffusion'' term arising from D, as well as the closely related test particle calculations, but is shown to have an important impact on the predicted fluxes. 16 refs., 1 fig

  14. 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

  15. 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.

  16. 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

  17. Recent advancement in research and planning toward high beta steady state operation in KSTAR

    International Nuclear Information System (INIS)

    The goal of Korean Superconducting Tokamak Advanced Research (KSTAR) research is to explore stable improved confinement regimes and technical challenge for superconducting tokamak operation and thus, to establish the basis for predictable high beta steady state tokamak plasma operation. To fulfil the goal, the current KSTAR research program is composed of three elements: 1) Exploration of anticipated engineering and technology for a stable long pulse operation of high beta plasmas including Edge Localized Mode (ELM) control with the low n (=1, 2) Resonant Magnetic Perturbation (RMP) using in-vessel control coils and innovative non-inductive current drives. The achieved long pulse operation up to ∼50s and fully non-inductive current drive will be combined in the future. Study of efficient heat exhaust will be combined with an innovative divertor design/operation. 2) Exploration of the operation boundary through establishment of true stability limits of the harmful MagnetoHydroDynamic (MHD) instabilities and confinement of the tokamak plasmas in KSTAR, making use of the lowest error field and magnetic ripple simultaneously achieved among all tokamaks ever built. The intrinsic machine error field has a long history of research as the source of MHD instabilities and magnetic ripple is known to be a cause of energy loss in the plasma. The achieved high beta discharges at βN ∼4 and stable discharges at q95 (∼2) will be further improved. 3) Validation of theoretical modeling of MHD instabilities and turbulence toward predictive capability of stable high beta plasmas. In support of these research goals, the state of the art diagnostic systems, such as Electron Cyclotron Emission Imaging (ECEI) system in addition to accurate profile diagnostics, are deployed not only to provide precise 2D/3D information of the MHD instabilities and turbulence but also to challenge unresolved physics problems such as the nature of ELMs, ELM-crash dynamics and the role of the core

  18. Observation of runaway electron beams by visible color camera in the Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    The synchrotron radiation originated from the energetic runaway electrons has been measured by a visible complementary metal oxide semiconductor camera working in the wavelength ranges of 380-750 nm in the Experimental Advanced Superconducting Tokamak [H. Q. Liu et al., Plasma Phys. Contr. Fusion 49, 995 (2007)]. With a tangential viewing into the plasma in the direction of electron approach on the equatorial plane, the synchrotron radiation from the energetic runaway electrons was measured in full poloidal cross section. The synchrotron radiation diagnostics provides a direct pattern of the runaway beam inside the plasma. The energy and pitch angle of runaway electrons have been obtained according to the synchrotron radiation pattern. A stable shell shape of synchrotron radiation has been observed in a few runaway discharges.

  19. 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

  20. Conceptual design of a fast-ion D-alpha diagnostic on experimental advanced superconducting tokamak.

    Science.gov (United States)

    Huang, J; Heidbrink, W W; Wan, B; von Hellermann, M G; Zhu, Y; Gao, W; Wu, C; Li, Y; Fu, J; Lyu, B; Yu, Y; Shi, Y; Ye, M; Hu, L; Hu, C

    2014-11-01

    To investigate the fast ion behavior, a fast ion D-alpha (FIDA) diagnostic system has been planned and is presently under development on Experimental Advanced Superconducting Tokamak. The greatest challenges for the design of a FIDA diagnostic are its extremely low intensity levels, which are usually significantly below the continuum radiation level and several orders of magnitude below the bulk-ion thermal charge-exchange feature. Moreover, an overlaying Motional Stark Effect (MSE) feature in exactly the same wavelength range can interfere. The simulation of spectra code is used here to guide the design and evaluate the diagnostic performance. The details for the parameters of design and hardware are presented. PMID:25430314

  1. Application of visible bremsstrahlung to Z(eff) measurement on the Experimental Advanced Superconducting Tokamak.

    Science.gov (United States)

    Chen, Yingjie; Wu, Zhenwei; Gao, Wei; Ti, Ang; Zhang, Ling; Jie, Yinxian; Zhang, Jizong; Huang, Juan; Xu, Zong; Zhao, Junyu

    2015-02-01

    The multi-channel visible bremsstrahlung measurement system has been developed on Experimental Advanced Superconducting Tokamak (EAST). In addition to providing effective ion charge Zeff as a routine diagnostic, this diagnostic can also be used to estimate other parameters. With the assumption that Zeff can be seen as constant across the radius and does not change significantly during steady state discharges, central electron temperature, averaged electron density, electron density profile, and plasma current density profile have been obtained based on the scaling of Zeff with electron density and the relations between Zeff and these parameters. The estimated results are in good coincidence with measured values, providing an effective and convenient method to estimate other plasma parameters. PMID:25725844

  2. Transient heat transport studies in JET conventional and advanced tokamak plasmas

    International Nuclear Information System (INIS)

    Transient transport studies are a valuable complement to steady-state analysis for the understanding of transport mechanisms and the validation of physics-based transport models. This paper presents results from transient heat transport experiments in JET and their modelling. Edge cold pulses and modulation of ICRH (in mode conversion scheme) have been used to provide detectable electron and ion temperature perturbations. The experiments have been performed in conventional L-mode plasmas or in Advanced Tokamak regimes, in the presence of an Internal Transport Barrier (ITB). In conventional plasmas, the issues of stiffness and non-locality have been addressed. Cold pulse propagation in ITB plasmas has provided useful insight into the physics of ITB formation. The use of edge perturbations for ITB triggering has been explored. Modelling of the experimental results has been performed using both empirical models and physics-based models. Results of cold pulse experiments in ITBs have also been compared with turbulence simulations. (author)

  3. Experimental Advanced Superconducting Tokamak/material and plasma evaluation system material migration experiment

    International Nuclear Information System (INIS)

    A test assembly of carbon coated molybdenum tiles was exposed to a sequence of well-diagnosed ohmic helium discharges in the Experimental Advanced Superconducting Tokamak (EAST) using the material and plasma evaluation system. The test assembly was configured to approximate, on a reduced scale, the profile of a first-wall panel in ITER. Net erosion and deposition of carbon was determined from measurements of the thickness of a carbon surface layer by ion-beam analysis before and after the plasma exposure. The highest erosion of 800 nm, or about one third of the initial thickness, occurred near the EAST midplane on the side facing the ion-drift direction. Erosion decreased in toroidal and poloidal directions with increasing distance from the plasma. Net deposition was not observed anywhere above the limit of detection. This experiment provides data needed to benchmark codes being used to predict erosion/deposition in ITER. (paper)

  4. CORBA-based solution for remote participation in SST-1 tokamak control and operation

    International Nuclear Information System (INIS)

    The steady state superconducting tokamak (SST-1) central control system is a distributed heterogeneous process communication system built on socket programming. It consists of machine, experiment and discharge control plus timing and a database. The software controls and monitors SST-1 subsystems: water-cooling, power supplies, cryogenics and vacuum over a local area network (LAN). The SST-1 control room is the place where all the activities like session announcement, machine control, experiment control, discharge control and monitoring are performed. We have realized that, instead of having a single monitoring place, we should have multiple monitoring points and it should be made possible to control the experiment from any PC over the LAN. In order to meet such requirements for remote participation in tokamak operation, we are upgrading the existing software. The upgraded software is based on Common Object Request Broker Architecture (CORBA) technology. The software is utilizing CORBA-services such as event service, naming services, interface repository and security services. The inherent features of CORBA make the software, platform and language independent. The software supports a variety of communication paradigms including publish-subscribe, peer-to-peer, and request-reply. Based on this software, one can participate in SST-1 tokamak operation from the LAN, or a wide area network (WAN) connection anywhere on the Internet. Each user can customize plasma parameters and diagnostics data that he wants to monitor, at any time without any change in the software and a copy of these parameters will be available to him. This paper focuses on the publish-subscribe communication paradigm and its application for a machine monitoring system

  5. Varennes Tokamak

    International Nuclear Information System (INIS)

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  6. 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.

  7. 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.

  8. 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.

  9. Study on lower hybrid current drive efficiency at high density towards long-pulse regimes in Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Significant progress on both L- and H-mode long-pulse discharges has been made recently in Experimental Advanced Superconducting Tokamak (EAST) with lower hybrid current drive (LHCD) [J. Li et al., Nature Phys. 9, 817 (2013) And B. N. Wan et al., Nucl. Fusion 53, 104006 (2013).]. In this paper, LHCD experiments at high density in L-mode plasmas have been investigated in order to explore possible methods of improving current drive (CD) efficiency, thus to extend the operational space in long-pulse and high performance plasma regime. It is observed that the normalized bremsstrahlung emission falls much more steeply than 1/ne-av (line-averaged density) above ne-av = 2.2 × 1019 m−3 indicating anomalous loss of CD efficiency. A large broadening of the operating line frequency (f = 2.45 GHz), measured by a radio frequency (RF) probe located outside the EAST vacuum vessel, is generally observed during high density cases, which is found to be one of the physical mechanisms resulting in the unfavorable CD efficiency. Collisional absorption of lower hybrid wave in the scrape off layer (SOL) may be another cause, but this assertion needs more experimental evidence and numerical analysis. It is found that plasmas with strong lithiation can improve CD efficiency largely, which should be benefited from the changes of edge parameters. In addition, several possible methods are proposed to recover good efficiency in future experiments for EAST

  10. 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

  11. 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

  12. 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.

  13. JT-60SA project for JA-EU broader approach satellite tokamak and national centralized tokamak

    International Nuclear Information System (INIS)

    JT-60 Super Advanced (JT-60SA) project is the joint project of ITER satellite tokamak by Japan and EU with Japanese Tokamak. The background, objects, device design, management of JT-60SA is stated. It consists of six chapters: the first chapter describes introduction, the second chapter states the objects of tokamak device complementing ITER, the third chapter contains research subjects and device performance such as plasma performance and demand for devices, operation scenario, control of MHD instability, and control of heat and particles, the forth chapter design of devices, the fifth chapter management and the sixth conclusion. In order to realize prototype reactor, improvement research of tokamak, development of reactor engineering technology, fusion reactor researches, tokamak theory and simulation, and social and environment safety research has to be advanced. (S.Y.)

  14. 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.

  15. 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

  16. Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

    International Nuclear Information System (INIS)

    The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DRI-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape control due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described

  17. 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

  18. First results on disruption mitigation by massive gas injection in Korea Superconducting Tokamak Advanced Research

    International Nuclear Information System (INIS)

    Massive gas injection (MGI) system was developed on Korea Superconducting Tokamak Advanced Research (KSTAR) in 2011 campaign for disruption studies. The MGI valve has a volume of 80 ml and maximum injection pressure of 50 bar, the diameter of valve orifice to vacuum vessel is 18.4 mm, the distance between MGI valve and plasma edge is ∼3.4 m. The MGI power supply employs a large capacitor of 1 mF with the maximum voltage of 3 kV, the valve can be opened in less than 0.1 ms, and the amount of MGI can be controlled by the imposed voltage. During KSTAR 2011 campaign, MGI disruptions are carried out by triggering MGI during the flat top of circular and limiter discharges with plasma current 400 kA and magnetic field 2–3.5 T, deuterium injection pressure 39.7 bar, and imposed voltage 1.1–1.4 kV. The results show that MGI could mitigate the heat load and prevent runaway electrons with proper MGI amount, and MGI penetration is deeper under higher amount of MGI or lower magnetic field. However, plasma start-up is difficult after some of D2 MGI disruptions due to the high deuterium retention and consequently strong outgassing of deuterium in next shot, special effort should be made to get successful plasma start-up after deuterium MGI under the graphite first wall.

  19. 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.

  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. 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.

  2. 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)

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

    International Nuclear Information System (INIS)

    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

  4. Analysis of pedestal gradient characteristic on the Experimental Advanced Superconducting Tokamak

    Science.gov (United States)

    Wang, Teng Fei; Han, Xiao Feng; Zang, Qing; Xiao, Shu Mei; Tian, Bao Gang; Hu, Ai Lan; Zhao, Jun Yu

    2016-05-01

    A pedestal database was built based on type I edge localized mode H-modes in the Experimental Advanced Superconducting Tokamak. The most common functional form hyperbolic tangent function (tanh) method is used to analyze pedestal characteristics. The pedestal gradient scales linearly with its pedestal top and the normalized pedestal pressure gradient α shows a strong correlation with electron collisionality. The connection among pedestal top value, gradient, and width is established with the normalized pedestal pressure gradient. In the core region of the plasma, the nature of the electron temperature stiffness reflects a proportionality between core and pedestal temperature while the increase proportion is lower than that expected in the high temperature region. However, temperature profile stiffness is limited or even disappears at the edge of the plasma, while the gradient length ratio ( ηe ) on the pedestal is important. The range of ηe is from 0.5 to 2, varying with the plasma parameters. The pedestal temperature brings a more significant impact on ηe than pedestal density.

  5. 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. PMID:26724032

  6. Advanced Control Test Operation (ACTO) facility

    International Nuclear Information System (INIS)

    The Advanced Control Test Operation (ACTO) project, sponsored by the US Department of Energy (DOE), is being developed to enable the latest modern technology, automation, and advanced control methods to be incorporated into nuclear power plants. The facility is proposed as a national multi-user center for advanced control development and testing to be completed in 1991. The facility will support a wide variety of reactor concepts, and will be used by researchers from Oak Ridge National Laboratory (ORNL), plus scientists and engineers from industry, other national laboratories, universities, and utilities. ACTO will also include telecommunication facilities for remote users

  7. 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

  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. 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

  10. Recent progress towards steady state tokamak operation with improved confinement in JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, Takaaki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2000-07-01

    In the JT-60U tokamak, optimization of high {beta}{sub p} mode and reversed shear mode plasmas are being done for establishment of scientific basis for steady state operation of tokamaks. In high {beta}{sub p} H-mode plasmas, {beta}{sub N}=2.9 and H{sub 89}=2.2-2.4 were sustained sanitarily by using high triangularity configuration and pressure profile optimization. Steady state performance was limited by resistive low toroidal mode number instabilities. Stabilization of resistive modes by using a newly installed ECRF system was attempted and a decrease of mode amplitude was observed but complete stabilization could not be achieved. In reversed shear plasmas, high fusion performance with equivalent DT fusion power gain of 0.5 was sustained for 0.8 s or an energy confinement time. The duration was limited by disruptive beta collapse that was encountered when the minimum value of q became 2 even with moderate beta, {beta}{sub N} {approx}1.2. Stationary sustainment of ITB was demonstrated in a full CD reversed shear plasma with LHCD. The sustainment of reversed shear current profile by bootstrap current was demonstrated in an ELMy H-mode edge reversed shear plasma with a high triangularity in a high q regime. A confinement enhancement factor of 3.5 and {beta}{sub N} of 2 were sustained for 2.7 s with stationary current and pressure profiles. Ar puffing to H mode plasmas aiming at high confinement with high density and high radiation fraction was performed and H{sub 89}{approx}1.4 with radiation fraction of 80% was obtained at 70% of Greenwald density. (author)

  11. 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

  12. 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

  13. 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.

  14. Operational efficiency subpanel advanced mission control

    Science.gov (United States)

    Friedland, Peter

    1990-01-01

    Herein, the term mission control will be taken quite broadly to include both ground and space based operations as well as the information infrastructure necessary to support such operations. Three major technology areas related to advanced mission control are examined: (1) Intelligent Assistance for Ground-Based Mission Controllers and Space-Based Crews; (2) Autonomous Onboard Monitoring, Control and Fault Detection Isolation and Reconfiguration; and (3) Dynamic Corporate Memory Acquired, Maintained, and Utilized During the Entire Vehicle Life Cycle. The current state of the art space operations are surveyed both within NASA and externally for each of the three technology areas and major objectives are discussed from a user point of view for technology development. Ongoing NASA and other governmental programs are described. An analysis of major research issues and current holes in the program are provided. Several recommendations are presented for enhancing the technology development and insertion process to create advanced mission control environments.

  15. 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)

  16. 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

  17. 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.

  18. The use of Computer Aided Management of Emergency Operations (CAMEO) at the Tokamak Fusion Test Reactor (TFTR)

    International Nuclear Information System (INIS)

    Computer Aided Management of Emergency Operations (CAMEO) is a software package for personal computers designed to enhance emergency preparedness and response involving chemical fires or spills. It was specifically designed by National Oceanic and Atmospheric Administration (NOAA) and the Environmental Protection Agency (EPA). The Princeton Plasma Physics Laboratory (PPPL) has combined fire fighting and hazmat response together in CAMEO and adapted it for use at the Tokamak Fusion Test Reactor (TFTR). 4 refs., 6 figs

  19. 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

  20. 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.)

  1. 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

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

    International Nuclear Information System (INIS)

    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

  3. 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

  4. 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

  5. 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.

  6. Design and exploitation of Advanced Tokamak scenarios at JET using the new Extreme Shape Controller

    International Nuclear Information System (INIS)

    In 2005 the JET machine entered a new exploitation phase with the MK2-HD (High Delta ) divertor, designed for ITER-relevant high triangularity and high elongation plasmas. The need to control the plasma boundary in these extreme equilibria with large variations of plasma parameters motivated the upgrade of the JET Shape Controller into the new fully model-based Extreme Shape Controller (XSC) [F. Sartori et al., Fusion Eng. and Design 74 (2005) 587-591, R. Albanese et al., Nucl. Fusion 38 (1998) 723-738]. An early version of XSC was tested in 2004 [F. Sartori et al., Proc of 31st EPS Conf. on Contr. Fusion and Plasma Physics, London 2004]; after minor modifications a new version has now been implemented. In this paper the use of the XSC tools for designing the complete time evolution of a physics discharge and its experimental validation will be presented. As test case, a JET Advanced Tokamak scenario at high triangularity [F.G. Rimini et al., Nucl. Fusion 45 (2005) 1481-1492] has been chosen. This scenario features a) the early formation of a single-null X-point configuration, b) the use of non-inductive Current Drive tools to tailor the current profile and c) the application of full power additional heating before the plasma current flat-top has been reached. This complex time evolution, with large variations of magnetic pressure, b, and internal inductance li, makes this scenario perfect for testing the XSC capabilities. Firstly the whole scenario is designed, using the equilibrium codes in the XSC tools suite, evolving from the early low triangularity to a high triangularity, d ∼ 0.4, equilibrium at the current flat-top. Next is the design of the actual controller, i.e. the choice of weights for the various poloidal field currents and geometric descriptors. In order to facilitate the boundary control different '' controllers '', i.e. sets of gains, have been devised for the low and the high triangularity phases. A closed-loop simulation of the whole scenario

  7. 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)

  8. 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)

  9. 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)

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

    Science.gov (United States)

    Zhang, Ling; Morita, Shigeru; Xu, Zong; Wu, Zhenwei; Zhang, Pengfei; Wu, Chengrui; Gao, Wei; Ohishi, Tetsutarou; Goto, Motoshi; Shen, Junsong; Chen, Yingjie; Liu, Xiang; Wang, Yumin; Dong, Chunfeng; Zhang, Hongmin; Huang, Xianli; Gong, Xianzu; Hu, Liqun; Chen, Junlin; Zhang, Xiaodong; Wan, Baonian; Li, Jiangang

    2015-12-01

    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(2) and pixel numbers of 1024 × 255 (26 × 26 μm(2)/pixel) is used for recording the focal image of spectral lines. An excellent spectral resolution of Δλ0 = 3-4 pixels, where Δλ0 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. PMID:26724029

  11. 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.

  12. 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.

  13. The Advanced Stellar Compass, Development and Operations

    DEFF Research Database (Denmark)

    Jørgensen, John Leif; Liebe, Carl Christian

    1996-01-01

    The science objective of the Danish Geomagnetic Research Satellite "Ørsted" is to map the magnetic field of the Earth, with a vector precision of a fraction of a nanotesla. This necessitates an attitude reference instrument with a precision of a few arcseconds onboard the satellite. To meet...... 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...

  14. The Advanced Stellar Compass, Development and Operations

    DEFF Research Database (Denmark)

    Jørgensen, John Leif; Liebe, Carl Christian

    The science objective of the Danish Geomagnetic Research Satellite "Ørsted" is to map the magnetic field of the Earth, with a vector precision of a fraction of a nanotesla. This necessitates an attitude reference instrument with a precision of a few arcseconds onboard the satellite. To meet 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 is...

  15. 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

  16. 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

  17. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. 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 of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  18. 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.

  19. Conceptual design of Fusion Experimental Reactor (FER) based on an advanced scenario of plasma operation and control

    International Nuclear Information System (INIS)

    The Fusion Experimental Reactor (FER) which is being developed at JAERI as a next-generation tokamak following JT-60 has the major purpose of realizing a self-ignited, long-burning DT plasma and demonstrating engineering feasibility. The paper emphasizes the advanced scenario of FER plasma operation and control and the advantage in engineering design made possible by the scenario. The FER concept is discussed, which is based on quasi-steady-state operation by a lower-hybrid-wave current drive or steady-state operation by three candidate radiofrequency waves, impurity control by a cold and dense divertor plasma and vertical position control of a highly elongated plasma. (author)

  20. Advances in the Electrical , Control Systems, General Analysis of the Coils Design in the Mexican Tokamak Experimental Facility

    International Nuclear Information System (INIS)

    Full text: The Fusion Research Group of the Autonomous University of Nuevo Leon (UANL, Spanish acronyms) presents its advances into the electrical, control systems and its coils design developed toward the Tokamak Experimental Facility [1]. This Research and Development Project (R+D) was approved from the Mexican Education Ministry (UANL-EXB-156). The present electrical and control systems studies are mainly oriented to establish our Magnetic Confinement Facility into our University Campus, with a D-shaped tokamak design with the next main characteristics: major radius R = 41 cm, minor radius a = 18.5 cm, aspect ratio A = 2.2162, safety factor q = 1.9552, plasma current Ip = 277 kA , toroidal field Bt = 1.3 T, electronic plasma density ne = 2 — 3 x 1013 cm-3. The present study at this time is an effective electrical engineering proposal to our University involving studies over the electrical power quality provided by Federal Electricity Commission. We define our parameters in voltage, current, frequency, to implement the correct strategies of electrical supplies in order to protect our facility. The analysis was performed measuring in the five domestic circuits of the University Campus: phase imbalances (current and voltage), harmonic distortion total and individual (voltages and currents of 1 — 50), transient capture, presentation of the power factor, registration of electrical interruptions and reclosing, measuring and recording quality power systems, crest factors (voltage and current), accurate RMS measurements of voltage and current, presentation of phasor diagrams. Our tokamak design contains a proposal coils arrangement capable for generate 1.6 T, with a coil current range (10,000 — 30, 000 A), short circuit times from 0.3 s to 1 s. The entirely systems uses Cu like first analysis material. The coils are designed with 3D CAD modeling and after, we apply finite element analysis through the software COMSOL Multiphysics. Our numerical calculus programs

  1. High Internal Inductance for High βN Steady-State Tokamak Operation

    Science.gov (United States)

    Ferron, J. R.

    2015-11-01

    An attractive scenario for steady-state tokamak operation at relatively high values of the internal inductance, li > 1 , has been demonstrated at DIII-D. The more peaked current density profile leads to reduced core energy transport and higher ideal stability limits that could eliminate the need for n >= 1 active stabilization coils at βN ~ 4, or enable βN ~ 5 with wall stabilization. The scenario's potential is shown by discharges at li ~ 1.3 with high bootstrap current fraction fBS ~0.8 , high plasma pressure βN ~ 5 and excellent confinement H98 (y , 2) ~ 1.8. This very high βN discharge with q95 =7.5 has noninductive current fraction fNI > 1 and too much bootstrap current in the H-mode pedestal, so li decreases with time. To achieve a stationary current profile, the key is to maximize βN and fBS while maintaining li high enough for stability through choice of q95 or by reduced pedestal current. DIII-D modeling shows that with q95 reduced to lower fBS to ~ 0.5, a self-consistent equilibrium has li ~ 1.07 and βN ~ 4 (below the n=1 no-wall limit) with q95 ~ 6. The remainder of the current can be externally-driven near the axis where the efficiency is high. Discharge tests with similar li in the ITER shape at q95=4.8 have reached fNI=0.7, fBS=0.4 at βN ~ 3.5 with performance appropriate for the ITER Q=5 mission, H89βN /q952~ 0.3. The li was shown to increase further above 1, to enable higher self-consistent fBS and βN, by reducing pedestal pressure and bootstrap current density through application of n = 3 resonant magnetic fields. With similar fields for ELM mitigation, and neutral beam and electron cyclotron current drive sources for near-axis current drive, the high li scenario is a potential option for ITER. The increased core confinement can help mitigate the effect of reduced pedestal pressure. Supported by US DOE under DE-FC02-04ER54698.

  2. Physics and control of ELMing H-mode negative-central-shear advanced tokamak ITER scenario based on experimental profiles from DIII-D

    International Nuclear Information System (INIS)

    Key DIII-D advanced tokamak (AT) experimental and modelling results are applied to examine the physics and control issues for ITER to operate in a negative central shear (NCS) AT scenario. The effects of a finite edge pressure pedestal and current density are included based on the DIII-D experimental profiles. Ideal and resistive stability analyses demonstrate that feedback control of resistive wall modes by rotational drive or flux conserving intelligent coils is crucial for these AT configurations to operate at attractive βN values in the range 3.0-3.5. Vertical stability and halo current analyses show that reliable disruption mitigation is essential and mitigation control using an impurity gas can significantly reduce the local mechanical stress to an acceptable level. Core transport and turbulence analyses indicate that control of the rotational shear profile is essential to reduce the pedestal temperature required for high β. Consideration of edge stability and core transport suggests that a sufficiently wide pedestal is necessary for the projected fusion performance. Heat flux analyses indicate that, with core-only radiation enhancement, the outboard peak divertor heat load is near the design limit of 10 MW m-2. Detached operation may be necessary to reduce the heat flux to a more manageable level. Evaluation of the ITER pulse length using a local step response approach indicates that the 3000 s ITER long-pulse scenario is probably both necessary and sufficient for demonstration of local current profile control. (author)

  3. 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.

  4. Advanced control of the Tokamak plasma shape and position by the quick response power supply

    International Nuclear Information System (INIS)

    The research on large tokamaks to get high parameter plasmas has been greatly extended. However, a number of engineering problems such as plasma vertical instability and unexpected pulse termination are still serious. For this reason the control of poloidal field power supplies employed to maintain the plasma in stable equilibrium with complex X points around plasma have been more and more important. Hybrid matrix control of shape and position for changing plasmas and development of the quick response power amplifier are reported. (author). 2 refs.; 6 figs.; 1 tab

  5. 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

  6. First results obtained from the soft x-ray pulse height analyzer on experimental advanced superconducting tokamak

    International Nuclear Information System (INIS)

    An assembly of soft x-ray pulse height analyzer system, based on silicon drift detector (SDD), has been successfully established on the experimental advanced superconducting tokamak (EAST) to measure the spectrum of soft x-ray emission (E=1-20 keV). The system, including one 15-channel SDD linear array, is installed on EAST horizontal port C. The time-resolved radial profiles of electron temperature and Kα intensities of metallic impurities have been obtained with a spatial resolution of around 7 cm during a single discharge. It was found that the electron temperatures derived from the system are in good agreement with the values from Thomson scattering measurements. The system can also be applied to the measurement of the long pulse discharge for EAST. The diagnostic system is introduced and some typical experimental results obtained from the system are also presented.

  7. 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. PMID:27131668

  8. Fast valve based on double-layer eddy-current repulsion for disruption mitigation in Experimental Advanced Superconducting Tokamak.

    Science.gov (United States)

    Zhuang, H D; Zhang, X D

    2015-05-01

    A fast valve based on the double-layer eddy-current repulsion mechanism has been developed on Experimental Advanced Superconducting Tokamak (EAST). In addition to a double-layer eddy-current coil, a preload system was added to improve the security of the valve, whereby the valve opens more quickly and the open-valve time becomes shorter, making it much safer than before. In this contribution, testing platforms, open-valve characteristics, and throughput of the fast valve are discussed. Tests revealed that by choosing appropriate parameters the valve opened within 0.15 ms, and open-valve times were no longer than 2 ms. By adjusting working parameter values, the maximum number of particles injected during this open-valve time was estimated at 7 × 10(22). The fast valve will become a useful tool to further explore disruption mitigation experiments on EAST in 2015. PMID:26026520

  9. 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.

  10. Effect of gas puffing from different side on lower hybrid wave-plasma coupling in experimental advanced superconductive tokamak

    International Nuclear Information System (INIS)

    Effect of gas puffing from electron-side and ion-side on lower hybrid wave (LHW)-plasma is investigated in experimental advanced superconductive tokamak for the first time. Experimental results with different gas flow rates show that electron density at the grill is higher in the case of gas puffing from electron-side; consequently, a lower reflection coefficient is observed, suggesting better effect of puffing from electron-side on LHW-plasma. The difference in edge density between electron- and ion-side cases suggests that local ionization of puffed gas plays a dominant role in affecting the density at the grill due to different movement direction of ionized electrons and that part of gas has been locally ionized near the gas pipe before diffusing into the grill region. Such difference could be enlarged and important in ITER due to the improvement of plasma parameters and LHW power

  11. Hydrocarbon deposition in gaps of tungsten and graphite tiles in Experimental Advanced Superconducting Tokamak edge plasma parameters

    International Nuclear Information System (INIS)

    The three-dimensional (3D) Monte Carlo code PIC-EDDY has been utilized to investigate the mechanism of hydrocarbon deposition in gaps of tungsten tiles in the Experimental Advanced Superconducting Tokamak (EAST), where the sheath potential is calculated by the 2D in space and 3D in velocity particle-in-cell method. The calculated results for graphite tiles using the same method are also presented for comparison. Calculation results show that the amount of carbon deposited in the gaps of carbon tiles is three times larger than that in the gaps of tungsten tiles when the carbon particles from re-erosion on the top surface of monoblocks are taken into account. However, the deposition amount is found to be larger in the gaps of tungsten tiles at the same CH4 flux. When chemical sputtering becomes significant as carbon coverage on tungsten increases with exposure time, the deposition inside the gaps of tungsten tiles would be considerable. (author)

  12. 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.

  13. 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

  14. 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

  15. 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

  16. Recent Advances in the long pulse heating and current drive system for KSTAR

    International Nuclear Information System (INIS)

    The heating and current-drive systems of KSTAR tokamak have been developed to support long pulse, high β, advanced tokamak physics experiments. Key technologies relevant for high power and long-pulse operation are under development. Substantial progresses have been made in areas such as ion source, RF launchers, tuning components and high power supplies and they will make the advanced tokamak operation of the KSTAR be obtainable and maintained for long-pulse operating condition. (author)

  17. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Czech Academy of Sciences Publication Activity Database

    Svoboda, V.; Kocman, J.; Grover, O.; Krbec, Jaroslav; Stöckel, Jan

    96-97, October (2015), s. 974-979. ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] Institutional support: RVO:61389021 Keywords : tokamak technology * remote participation * plasma stabilization Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://dx.doi.org/10.1016/j.fusengdes.2015.06.044

  18. An operational test of a time-of-flight analyser at the ASDEX tokamak

    International Nuclear Information System (INIS)

    The performance of a time-of-flight energy analyser, used for the investigation of the slowing-down spectrum of the heating beams at the ASDEX tokamak, is described. The time-of-flight analyser has a short flight path (15 cm). Its energy resolution amounts to a few per cent. The analyser was equipped with a preselecting, achromatic magnet system to separate the fast neutrals from thermal plasma particles and light

  19. Operational start-up and vessel conditioning of the Novillo tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Melendez-Lugo, L.; Lopez-Callejas, R.; Valencia-Alvarado, R.; Chavez-Alarcon, E.; Colunga-Sanchez, S.; Jimenez-Dominguez, H.; Gaytan-Gallardo, E. (Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico))

    1992-01-01

    The first results of design, construction and stage of the Novillo Tokamak are presented. The main parameters of the device are: R[sub o]=0.23m, a=0.08m, a[sub p]=0.06m, q[approx]3, B[sub 1][approx]0.5T, B[sub p]=0.04T, I[sub p]=12kA, n[sub a]=2x10[sup 19]m[sup -3], T[sub a]=150eV, T[sub i]=50eV.The vacuum vessel has a wide accessibility to all major forms of diagnostics covering approximately 25% of the total area. We have obtained tokamak discharges with base pressure as low as 1.0x10[sup -8] Torr after a vessel conditioning. Using approximately 70% of OHT energy, the plasma current has been more than 11 kA. At the current stage of the Novillo project, plasma measurements have been restricted to magnetic probes and a more sophisticated diagnostic like a 2mm microwave interferometer and a 9 mm reflectometer and now a 32 channel acquisition data system are setting on the tokamak and will be available soon. (author) 3 refs., 3 figs., 1 tab.

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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 co-ordinated 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 Co-ordinated Research Project is presented. (author)

  5. Advanced methods in global gyrokinetic full f particle simulation of tokamak transport

    International Nuclear Information System (INIS)

    A new full f nonlinear gyrokinetic simulation code, named ELMFIRE, has been developed for simulating transport phenomena in tokamak plasmas. The code is based on a gyrokinetic particle-in-cell algorithm, which can consider electrons and ions jointly or separately, as well as arbitrary impurities. The implicit treatment of the ion polarization drift and the use of full f methods allow for simulations of strongly perturbed plasmas including wide orbit effects, steep gradients and rapid dynamic changes. This article presents in more detail the algorithms incorporated into ELMFIRE, as well as benchmarking comparisons to both neoclassical theory and other codes.Code ELMFIRE calculates plasma dynamics by following the evolution of a number of sample particles. Because of using an stochastic algorithm its results are influenced by statistical noise. The effect of noise on relevant magnitudes is analyzed.Turbulence spectra of FT-2 plasma has been calculated with ELMFIRE, obtaining results consistent with experimental data

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

    International Nuclear Information System (INIS)

    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.)

  7. 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.)

  8. 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

  9. 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

  10. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li2O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  11. Edge Thomson scattering diagnostic on COMPASS tokamak: Installation, calibration,operation, improvements

    Czech Academy of Sciences Publication Activity Database

    Böhm, Petr; Aftanas, Milan; Bílková, Petra; Štefániková, Estera; Mikulín, Ondřej; Melich, Radek; Janky, Filip; Havlíček, Josef; Šesták, David; Weinzettl, Vladimír; Stöckel, Jan; Hron, Martin; Pánek, Radomír; Scannell, R.; Frassinetti, L.; Fassina, A.; Naylor, G.; Walsh, M.J.

    2014-01-01

    Roč. 85, č. 11 (2014), 11E431-11E431. ISSN 0034-6748. [Topical Conference on High-Temperature Plasma Diagnostics/20./. Atlanta, Georgia, 01.06.2014-05.06.2014] R&D Projects: GA MŠk(CZ) LM2011021; GA ČR(CZ) GA14-35260S Institutional support: RVO:61389021 Keywords : plasma * tokamak * pedestal * Thomson scattering * diagnostic Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.614, year: 2014 http://scitation.aip.org/content/aip/journal/rsi/85/11/10.1063/1.4893995

  12. Personnel protection during the operation of Thomson scattering laser system on COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Böhm, Petr; Hron, Martin; Kovar, J.; Sova, J.; Zvolanek, M.; Aftanas, Milan; Bílková, Petra; Pánek, Radomír; Walsh, M.J.

    2011-01-01

    Roč. 86, 6-8 (2011), s. 699-702. ISSN 0920-3796. [Symposium on Fusion Technology, SOFT-26/26th./. Porto, 27.09.2010-01.10.2010] R&D Projects: GA ČR GA202/09/1467; GA ČR GD202/08/H057; GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * Thomson scattering * Laser safety * Personnel protection * PLC Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.490, year: 2011 http://www.sciencedirect.com/science/article/pii/S0920379611002432

  13. 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.

  14. 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

  15. Recent advances in the theory and simulation of pellet ablation and fast fuel relocation in tokamaks

    International Nuclear Information System (INIS)

    Full text: This paper presents new theory and simulation of pellet ablation, and the rapid cross-field relocation ionized pellet substance following HFS pellet injection in tokamaks. A unique time-dependent 2-D Eulerian code CAP, was developed that is the first to self- consistently treat the key B-field effects: (1) Pellet deformation into a 'pancake' shape, driven by the anisotropic surface ablation pressure, can cut pellet lifetimes by almost ∼3x (2) J x B funneling of the flow into a field-aligned cigar-shaped structure enhances shielding. Near-pellet cloud parameters from CAP are critical inputs for PRL and AMR codes that model fast advection of the 'detached' clouds accelerated by the ∇B effect. PRL contains new geometrical effects of toroidicity, magnetic shear, and curvature drifts by parallel flows. Consequently, the calculated fuel deposition is in better accord with density measurements on DIII-D, providing improved predictive capability for ITER. A new 3-D MHD simulation code AMR can provide the required fine-scale mesh size needed for accurate modeling of strongly localized pellet clouds. (author)

  16. Next-generation plasma control in the DIII-D tokamak

    International Nuclear Information System (INIS)

    The advanced tokamak (AT) operating mode, which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. This paper describes progress towards the DIII-D AT mission goal through both improvements in real-time computational hardware and control algorithm capability. A number of device constraints, some unique to DIII-D, and their impact on operational shape and position control are discussed. Some partial solutions are described

  17. 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

  18. A need for non-tokamak approaches to magnetic fusion energy

    International Nuclear Information System (INIS)

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. Y.; Hong, J. H.; Jang, J. H.; Park, J. S.; Choe, Wonho, E-mail: wchoe@kaist.ac.kr [Department of Physics, Korea Advanced Institute of Science and Technology (KAIST), 34141 Daejeon (Korea, Republic of); Impurity and Edge Plasma Research Center, KAIST, 34141 Daejeon (Korea, Republic of); Hahn, S. H.; Bak, J. G.; Lee, J. H.; Ko, W. H.; Lee, K. D.; Lee, S. H.; Lee, H. H.; Juhn, J.-W.; Kim, H. S.; Yoon, S. W.; Han, H. [National Fusion Research Institute, 34133 Daejeon (Korea, Republic of); Ghim, Y.-C. [Deparment of Nuclear and Quantum Engineering, KAIST, 34141 Daejeon (Korea, Republic of)

    2015-12-15

    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 D{sub α} 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.

  20. 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)

  1. Availability of ITER and Fusion Power Plants considering the experience of the present tokamaks and ITER operational constraints

    International Nuclear Information System (INIS)

    ITER represents a very important step forward in nuclear fusion field from both physics and technology point of view towards Fusion Power Plants (FPPs). Its availability is vital to allow the accomplishment of the planned experimental programmes. On the other side ITER is a first kind of nuclear machine with respect to size and operational and safety issues. Present tokamaks are prototypical as well. Their operating experience has now reached several hundred device-years and gives useful indications for the devices themselves and for those under design or construction. Most of the results are expressed in terms of good plasma pulse number, operating time and delays, effective operation time versus planned operation time. Thanks to the learning process, the availability of present fusion machines, mainly affected by reliability (less from maintainability) has been maintained constant and in some case improved all over the years in spite of the ageing of systems and components and of more demanding and complex plasma scenarios. ITER (and FPPs) will face further challenges, like more significant steady state and transient loads, much higher neutronic flux and fluence, activation and dose rates, that will ask for a periodic replacement of important components with remote handling techniques as well as the control of safety operating limits and conditions (OLCs), like inventories of tritium and dust. The strict quality requirements on components and maintenance equipment for safety and investment protection could compensate partially the reduction of availability. The paper will present in a qualitative way the main factors affecting the availability of ITER, extrapolating also availability data and the lessons learnt from JET, JT-60, DIII-D, ASDEX-U, TORE SUPRA, etc. It will report also reflections about the areas where to concentrate more efforts in order to optimise the availability considering the main constraints present in the design and in the operation

  2. Limits of possible operation of the R-tokamak due to ideal MHD instabilities

    International Nuclear Information System (INIS)

    A series of MHD stability calculations has been made with the ERATO code to know the maximum β which can be expected for the R-Tokamak using the parameters of the second phase design (an aspect ratio of 2.75, ellipticity of 1.8 and triangularity of 0.3). The highest β obtained under the condition that both the n=1 free boundary mode (with no wall stabilization) and the n=infinity ballooning modes are stable is about 6% at qsub(s)--2.0 and qsub(o) at the Mercier limit on axis. This result has been found by performing some optimization of the current and pressure profiles. If only the ballooning modes are concerned, the limiting β becomes 8.6%. The sensitivity of the result to elongation has been studied. It has been found that the maximum β increases and then decreases with elongation. The optimum β is obtained for an elongation of 1.8 if both the n=1 kink and n=infinity ballooning limits are considered and of 2.0 if only ballooning modes are considered. These results are compared with proposed scaling laws. (author)

  3. 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

  4. 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

  5. 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.

  6. JINTRAC. A system of codes for integrated simulation of tokamak scenarios

    International Nuclear Information System (INIS)

    Operation and exploitation of present and future Tokamak reactors require advanced scenario modeling in order to optimize engineering parameters in the design phase as well as physics performance during the exploitation phase. The simulation of Tokamak scenarios involves simultaneous modeling of different regions of the reactor, characterized by different physics and symmetries, in order to predict quantities such as particle and energy confinement, fusion yield, power deposited on wall, wall load from fast particles. JINTRAC is a system of 25 interfaced Tokamak-physics codes for the integrated simulation of all phases of a Tokamak scenario. JINTRAC predictions reflect the physics and assumptions implemented in each module and extensive comparison with experimental data is needed to allow validation of the models and improvement of Tokamak-physics understanding. (author)

  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. Neutron production and detection on the TFTR tokamak during DT operation

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) has performed initial high-power experiments with deuterium and tritium fueling at nominally equal densities. Peak fusion power production of 6.2 ± 0.5 MW has been achieved in plasmas heated for 0.7 seconds by 29.5 MW of similar powers of deuterium and tritium neutral beams. The fusion power production and associated alpha-particle density and pressure from the d(t,n)α reactions are determined by a set of neutron diagnostics with complementary energy, time, and space resolution. These 2.0 MA plasma current, 5.1 T toroidal magnetic field, 2.52 m major radius and 0.87 m minor radius discharges have an inferred central alpha particle density of 1.2 x 1017 m-3 and central alpha particle pressure Pα corresponding to βα = 2μoPα/B2 of 2.2 x 10-3. During the deuterium-tritium experiments, neutron emission rates and yields were measured with 235U and 238U fission chambers, silicon surface barrier diodes, a scintillating fiber detector, spatially collimated 4He proportional counters and ZnS scintillators, and a variety of elemental activation foils. The activation foils, 4He counters, scintillating fibers, and silicon diodes can discriminate between 14-MeV and 2.5-MeV neutrons by nuclear reaction thresholds or pulse height discrimination. The other detectors are sensitive to both d(d,n)3He and dt neutrons, and are generally more sensitive to the latter. Use of the 115In(n,n') 115mIn activation reaction in conjunction with pure-DT reactions allows a determination of the DT/(DD + DT) ratio in trace tritium or low-power tritium beam experiments. The ZnS scintillator detectors and 4He proportional counters are located in a 10-channel neutron collimator and provide spatial imaging of the neutron emission

  9. 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...

  10. Design and construction of the KSTAR tokamak

    International Nuclear Information System (INIS)

    The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and 'advanced tokamak' physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long-pulse current-drive and heating systems. The 'advanced tokamak' aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. (author)

  11. 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.)

  12. 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.

  13. Steady-state operation of tokamaks: Key physics and technology developments on Tore Supra

    International Nuclear Information System (INIS)

    Important technological and physics issues related to long pulse operation required for a reactor are now being addressed in Tore Supra. experimental results in conditions where all the plasma facing components are actively cooled during pulses exceeding six minutes. Important physics issues related to continuous operation are observed in non inductively driven plasmas. (author)

  14. NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    OAK A271 NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK. The advanced tokamak (AT) operating mode which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. Simultaneous high performance regulation of the plasma boundary and internal profiles requires multivariable control techniques to account for the highly coupled influences of equilibrium shape, profile, and stability control. This paper describes progress towards the DIII-D At mission goal through both significantly improved real-time computational hardware and control algorithm capability

  15. 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

  16. 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)

  17. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Full text: SST-1, a steady state superconducting tokamak, is at advanced stage of erection at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation and triangularity. The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 T at plasma center and a plasma current of 220 kA. Hydrogen gas will be used and plasma discharge duration will be 1000 s. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors having ports and 16 rings with D- shaped cross-section, which are welded in-situ during the SST-1 assembly. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as Sc magnets and cryostat, to minimize the radiation losses at the Sc magnets. In SST-1 tokamak, the auxiliary current drive will be based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. The assembly of the SST-1 tokamak is nearing completion. The cool down of the Superconducting magnets is scheduled to start by middle of year 2004

  18. 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

  19. Development of advanced tokamak scenarios based on high bootstrap currents in JET

    International Nuclear Information System (INIS)

    High bootstrap current experiments with the bootstrap fraction (Ibootstrap/Iplasma) up to 0.7 at 1 MA and 0.5 at 1.5 MA were previously achieved in JET in plasmas with high q(a), high triangularity and high confinement (H≥3). During initial operation with the new JET pumped divertor, the domain of parameters has been extended to cover reactor relevant domains such as low q(a) and high βn; up to βn = 3. High beta poloidal plasmas (necessary for the achievement of high bootstrap fractions) with βp up to 1 for 2 MA plasmas and with βp up to 2 for 1 MA plasmas have been achieved. Values of βp = 1.5 and βn = 3 have been obtained simultaneously over several seconds at q95 4.5 The confinement of these plasmas is lower than in the previous campaign but these discharges display 'quasi steady state' characteristic. In order to improve the confinement of these discharges, several parameters have been varied such as plasma volume, triangularity and βp itself, showing no significant benefit. Configurations using non-inductive current drive to produce stable higher confinement plasmas are being developed. A 'deep' shear reversal configuration has been established, and initial data are presented. (author). 5 refs, 7 figs

  20. 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.

  1. Elements of an advanced integrated operator control station

    International Nuclear Information System (INIS)

    One of the critical determinants of performance for any remotely operated maintenance system is the compatibility achieved between elements of the man/machine interface (e.g., master manipulator controller, controls, displays) 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

  2. 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

  3. Elements of an advanced integrated operator control station

    International Nuclear Information System (INIS)

    One of the critical determinants of performance for any remotely operated maintenance system is the compatibility achieved between elements of the man/machine interface (e.g., master manipulator controller, controls, displays) 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

  4. 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.

  5. 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)

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

    International Nuclear Information System (INIS)

    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

  7. Advances in new WWER designs to improve operation and maintenance

    International Nuclear Information System (INIS)

    Economic operational indices of WWER-type reactors show their competitiveness in all the countries where these reactors operate. Advanced WWERs being designed and constructed now have the improved characteristics of economical efficiency and are more convenient for operation and maintenance. Many technical solutions aimed at improvement of the operational performance are implemented in the design of WWER-1000/V-392 and WWER-640/V-407, and these reactors are the important basis for the nuclear power expansion in Russia. Some of these solutions are considered in the present paper. (author)

  8. 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

  9. Bootstrap current fraction scaling for a tokamak reactor design study

    International Nuclear Information System (INIS)

    Highlights: • New bootstrap current fraction scalings for systems codes were derived by solving the Hirshman–Sigmar model. • Nine self-consistent MHD equilibria were constructed in order to compare the bootstrap current fraction values. • Wilson formula most successfully predicted the bootstrap current fraction, but it requires current density profile index. • The new scaling formulas and IPDG accurately estimated the fBS values for the normal and weakly reversed shear tokamaks. - Abstract: We have derived new bootstrap current fraction scalings for systems codes by solving the Hirshman–Sigmar model, which is valid for arbitrary aspect ratios and collision conditions. The bootstrap current density calculation module in the ACCOME code was used with the matrix inversion method without the large aspect ratio assumption. Nine self-consistent MHD equilibria, which cover conventional, advanced and spherical tokamaks with normal or reversed shear, were constructed using numerical calculations in order to compare the bootstrap current fraction values with those of the new model and all six existing models. The Wilson formula successfully predicted the bootstrap current fraction, but it requires current density profile index for the calculation. The new scaling formulas and IPDG accurately estimated the bootstrap current fraction for the normal and weakly reversed shear tokamaks, regardless of the aspect ratio. However, none of the existing models except the Wilson formula can accurately estimate the bootstrap current fraction for the reversed shear tokamaks, which is promising for the advanced tokamak operation mode

  10. Power supplies and quench protection for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems

  11. 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...

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. 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

  19. Bayesian derivation of plasma equilibrium distribution function for tokamak scenarios and the associated Landau collision operator

    Science.gov (United States)

    Di Troia, C.

    2015-11-01

    A class of parametric distribution functions was proposed in (Di Troia 2012 Plasma Phys. Control. Fusion 54 105017) as equilibrium distribution functions (EDFs) for charged particles in fusion plasmas, representing supra-thermal particles in anisotropic equilibria for Neutral Beam Injection and Ion Cyclotron Heating scenarios. Moreover, those EDFs can be used to represent also nearly isotropic equilibria for Slowing-Down alpha particles and core thermal plasma populations. Such EDFs depend on constants of motion (COMs). In axisymmetric system with no equilibrium electric field, they depend on toroidal canonical momentum {{P}φ} , kinetic energy w and magnetic moment μ. In the present work, the same EDFs are obtained from first principles and general hypothesis. The derivation is probabilistic and makes use of the Bayes’ Theorem. The bayesian argument is used to describe how the plasma is far from the prior probability distribution function (pdf), e.g. Maxwellian, based on the information obtained from magnetic moment and guiding center velocity pdf. Once the general functional form of the EDF has been settled, it is shown how to associate a modified Landau collision operator in the Fokker-Planck equation, to describe the system relaxation towards the proposed EDF.

  20. Problems and issues for tokamak long-pulse operation: major progress on Tore Supra

    International Nuclear Information System (INIS)

    This paper gives an overview of the major progress achieved recently, taking advantage of the new configuration of Tore-Supra. As a background, the main results obtained in the nineties are first recalled, as well as the identified limitations that led to the upgrade project. In the new configuration, the Tore Supra operational domain has experienced a rapid extension. Indeed, discharge of more than 4 minutes have been performed with a record breaking discharge accounting for 0.75 GJ of injected/extracted energy. Primary transformer recharge by Lower Hybrid Current Drive (LHCD) has been observed during 1 minute while fully steady state discharge, with complete diffusion of the resistive current, are performed routinely (typical parameters: plasma current 0.52 MA, toroidal magnetic field 4 T, LH power around 3 MW, electron line density 2.5 1019 m-2), limited in duration by the original LHCD system. Ion cyclotron waves (ICRH) have been coupled to plasma during one minutes in combination with LHCD, in a higher density scenarios (Greenwald fraction of 0.8, 0.11 GJ of injected ICRH power for 0.42 GJ total injected power) and with a significant fraction of bootstrap current (15-20%). Electron cyclotron current drive experiments are also carried out for current profile control and for the study of current drive synergy with LHCD. Opportunely, a new record of EC injected energy has been established in a single ECRH pulse of 32 s (25 MJ). In these discharges, stable central electron temperature oscillations sometimes appear, probably due to the interplay between heat transport and current drive. Density profile peaking is observed despite the absence of toroidal field, suggesting the existence of a turbulent inward pinch. Finally, article balance analysis indicate that the in-vessel deuterium inventory never reach saturation. found any carbon deposits and flakes have in the inner vessel, possibly playing a role in the retention. (author)

  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. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Monte Carlo simulation of a Bonner sphere spectrometer for application to the determination of neutron field in the Experimental Advanced Superconducting Tokamak experimental hall.

    Science.gov (United States)

    Hu, Z M; Xie, X F; Chen, Z J; Peng, X Y; Du, T F; Cui, Z Q; Ge, L J; Li, T; Yuan, X; Zhang, X; Hu, L Q; Zhong, G Q; Lin, S Y; Wan, B N; Gorini, G; Li, X Q; Zhang, G H; Chen, J X; Fan, T S

    2014-11-01

    To assess the neutron energy spectra and the neutron dose for different positions around the Experimental Advanced Superconducting Tokamak (EAST) device, a Bonner Sphere Spectrometer (BSS) was developed at Peking University, with totally nine polyethylene spheres and a SP9 (3)He counter. The response functions of the BSS were calculated by the Monte Carlo codes MCNP and GEANT4 with dedicated models, and good agreement was found between these two codes. A feasibility study was carried out with a simulated neutron energy spectrum around EAST, and the simulated "experimental" result of each sphere was obtained by calculating the response with MCNP, which used the simulated neutron energy spectrum as the input spectrum. With the deconvolution of the "experimental" measurement, the neutron energy spectrum was retrieved and compared with the preset one. Good consistence was found which offers confidence for the application of the BSS system for dose and spectrum measurements around a fusion device. PMID:25430324

  17. 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

  18. Design, construction, and first operational results of a 5 megawatt feedback controlled amplifier system for disruption control on the Columbia University HBT-EP tokamak

    International Nuclear Information System (INIS)

    This paper presents the electrical design and first operational results of a 5 Megawatt feedback controlled amplifier system designed to drive a 300 uH saddle coil set on the HBT-EP tokamak. It will be used to develop various plasma feedback techniques to control and inhibit the onset of plasma disruptions that are observed in high β plasmas. To provide a well characterized system, a high fidelity, high power closed loop amplifier system has been refurbished from the Los Alamos ZT-P equilibrium feedback system. In its configuration developed for the Columbia HBT-EP tokamak, any desired waveform may be generated within a 1,100 ampere and 16 kV peak to peak dynamic range. An energy storage capacitor bank presently limits the effective full power pulse width to 10 mS. The full power bandwidth driving the saddle coil st is ∼12 kHz, with bandwidth at reduced powers exceeding 30 kHz. The system is designed similar to a grounded cathode, push-pull, transformer coupled, tube type amplifier system. The detailed electrical design of the power amplifier, transformer, and feedback system will be provided in addition to recent HBT-EP operational results

  19. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  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. 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)

  2. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    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)

  3. 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

  4. Advanced operator interface design for CANDU-3 fuel handling system

    International Nuclear Information System (INIS)

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System'. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author)

  5. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  6. 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

  7. 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

  8. JT-60 power tests from mechanical and thermal viewpoints of tokamak machine

    International Nuclear Information System (INIS)

    JT-60 power tests were carried out, to demonstrate, in advance of actual plasma operation, satisfactory performance of the tokamak machine, power suppliers and control system in combination. The tests began with low power ones of individual coil systems, progressed to full power ones and concluded successfully. The present paper describes the principal results of JT-60 power tests from mechanical and thermal viewpoints of tokamak machine. All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and mechanical integrity of tokamak machine. Measured strain and displacement showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. Vibration of the vacuum vessel was found to be large up to 48 m/s/sup 2/ and caused excessive vibration of the lateral port gate-valves. A few limitations to machine operation were also made clear quantatively

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hill, W.E.; Houser, M.M.; Knee, H.E.; Spelt, P.F.

    1995-03-01

    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.

  11. 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

  12. 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)

  13. PSA-operations synergism for the advanced test reactor shutdown operations PSA

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) Probabilistic Safety Assessment (PSA) for shutdown operations, cask handling, and canal draining is a successful example of the importance of good PSA-operations synergism for achieving a realistic and accepted assessment of the risks and for achieving desired risk reduction and safety improvement in a best and cost-effective manner. The implementation of the agreed-upon upgrades and improvements resulted in the reductions of the estimated mean frequency for core or canal irradiated fuel uncovery events, a total reduction in risk by a factor of nearly 1000 to a very low and acceptable risk level for potentially severe events

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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)

  19. Current profile modeling to extend the duration of high performance advanced tokamak modes in DIII-D

    International Nuclear Information System (INIS)

    In DIII-D, as in a number of tokamaks, high performance is obtained with various optimized magnetic shear configurations that exhibit internal transport barriers. Negative central shear (NCS) discharges are created transiently during the current ramp-up by auxiliary heating and current drive from neutral beam injection. Both qmin and the radius at which it occurs, ρqmin, decrease with time as the Ohmic current diffuses inward. The q-profiles calculated using EFIT with external magnetic and Motional Stark Effect (MSE) measurements as constraints are comparable to those calculated with the Corsica code, a time-dependent, 2D equilibrium and 1D transport modeling code. Corsica is used to predict the temporal evolution of the current density from a combination of measured profiles, transport models and neoclassical resistivity. Using these predictive capabilities, the authors are exploring methods for increasing the duration and ρqmin of the NCS configuration by local control of the current density profile with simulations of the possible control available from the electron cyclotron heating and current drive system currently being upgraded on DIII-D. Their intention is not to do a detailed investigation of transport models but rather to provide a reasonable model of heat conductivity to be able to simulate effects of electron cyclotron heating (ECH) and current drive (ECCD) on confinement in NCS configurations. The authors adjust free parameters (c, c1 and c2) in the model to obtain a reasonable representation of the temporal evolution of electron and ion temperature profiles consistent with those measured in selected DIII-D shots. In all cases, they use the measured density profiles rather than self-consistently solve for particle sources and particle transport at this time

  20. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  1. 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

  2. Development of integrated real-time control of internal transport barriers in advanced operation scenarios on Jet

    International Nuclear Information System (INIS)

    An important experimental programme is in progress on JET to investigate plasma control schemes which, with a limited number of actuators, could eventually enable ITER to sustain steady state burning plasmas in an 'advanced tokamak' operation scenario. A multi-variable model-based technique was recently developed for the simultaneous control of several plasma parameter profiles in discharges with internal transport barriers (ITB), using lower hybrid current drive (LHCD) together with neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH). The proposed distributed-parameter control scheme relies on the experimental identification of an integral linear response model operator and retains the intrinsic couplings between the plasma parameter profiles. A first set of experiments was performed to control the current density profile in the low-density/low-power LH-driven phase of the JET advanced scenarios, using only one actuator (LHCD) and a simplified (lumped-parameter) version of the control scheme. Several requested steady state magnetic equilibria were thus obtained and sustained for about 7 s, up to full relaxation of the ohmic current throughout the plasma. A second set of experiments was dedicated to the control of the q-profile with 3 actuators (LHCD, NBI and ICRH) during the intense heating phase of advanced scenarios. The safety factor profile was also shown to approach a requested profile within about 5 s. The achieved plasma equilibrium was close to steady state. Finally, during the recent high power experimental campaign, experiments have been conducted in a 3 T / 1.7 MA plasma, achieving the simultaneous control of the current density and electron temperature profiles in ITB plasmas. Here, the distributed-parameter version of the algorithm was used for the first time, again with 3 actuators. Real-time control was applied during 7 s, and allowed to reach successfully different target q-profiles (monotonic and reversed-shear ones) and

  3. Development of integrated real-time control of internal transport barriers in advanced operation scenarios on Jet

    Energy Technology Data Exchange (ETDEWEB)

    Moreau, D.; Laborde, L.; Litaudon, X.; Mazon, D.; Zabeo, L.; Joffrin, E.; Lennholm, M. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Moreau, D. [EFDA-JET CSU, Culham Science Centre, Abingdon, OX (United Kingdom); Crisanti, F.; Pericoli-Ridolfini, V.; Riva, M.; Tuccillo, A. [Euratom-ENEA Association, C.R. Frascati (Italy); Murari, A. [Euratom-ENEA Association, Consorzio RFX, Padova (Italy); Tala, T. [Euratom-TEKES Association, VTT Processes (Finland); Albanese, R.; Ariola, M.; Tommasi, G. de; Pironti, A. [Euratom-ENEA Association, CREATE, Napoli (Italy); Felton, R.; Zastrow, K.D. [Euratom-UKAEA Association, Culham Science Centre, Abingdon(United Kingdom); Baar, M. de; Vries, P. de [Euratom-FOM Association, TEC Cluster, Nieuwegein (Netherlands); La Luna, E. de [Euratom-CIEMAT Association, CIEMAT, Madrid (Spain)

    2004-07-01

    An important experimental programme is in progress on JET to investigate plasma control schemes which, with a limited number of actuators, could eventually enable ITER to sustain steady state burning plasmas in an 'advanced tokamak' operation scenario. A multi-variable model-based technique was recently developed for the simultaneous control of several plasma parameter profiles in discharges with internal transport barriers (ITB), using lower hybrid current drive (LHCD) together with neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH). The proposed distributed-parameter control scheme relies on the experimental identification of an integral linear response model operator and retains the intrinsic couplings between the plasma parameter profiles. A first set of experiments was performed to control the current density profile in the low-density/low-power LH-driven phase of the JET advanced scenarios, using only one actuator (LHCD) and a simplified (lumped-parameter) version of the control scheme. Several requested steady state magnetic equilibria were thus obtained and sustained for about 7 s, up to full relaxation of the ohmic current throughout the plasma. A second set of experiments was dedicated to the control of the q-profile with 3 actuators (LHCD, NBI and ICRH) during the intense heating phase of advanced scenarios. The safety factor profile was also shown to approach a requested profile within about 5 s. The achieved plasma equilibrium was close to steady state. Finally, during the recent high power experimental campaign, experiments have been conducted in a 3 T / 1.7 MA plasma, achieving the simultaneous control of the current density and electron temperature profiles in ITB plasmas. Here, the distributed-parameter version of the algorithm was used for the first time, again with 3 actuators. Real-time control was applied during 7 s, and allowed to reach successfully different target q-profiles (monotonic and reversed-shear ones

  4. Enterprise SRS: Leveraging Ongoing Operations to Advance National Programs - 13108

    International Nuclear Information System (INIS)

    The SRS is re-purposing its vast array of assets to solve future national issues regarding environmental stewardship, national security, and clean energy. The vehicle for this transformation is Enterprise SRS which presents a new, strategic view of SRS as a united endeavor for 'all things nuclear' as opposed to a group of distinct and separate entities with individual missions and organizations. Key among the Enterprise SRS strategic initiatives is the integration of research into facilities in conjunction with ongoing missions to provide researchers from other national laboratories, academic institutions, and commercial entities the opportunity to demonstrate their technologies in a relevant environment and scale prior to deployment. To manage that integration of research demonstrations into site facilities, The DOE Savannah River Operations Office, Savannah River Nuclear Solutions, and the Savannah River National Laboratory (SRNL) have established the Center for Applied Nuclear Materials Processing and Engineering Research (CANMPER). The key objective of this initiative is to bridge the gap between promising transformational nuclear materials management advancements and large-scale deployment of the technology by leveraging SRS assets (e.g. facilities, staff, and property) for those critical engineering-scale demonstrations necessary to assure the successful deployment of new technologies. CANMPER will coordinate the demonstration of R and D technologies and serve as the interface between the engineering-scale demonstration and the R and D programs, essentially providing cradle-to-grave support to the R and D team during the demonstration. While the initial focus of CANMPER will be on the effective use of SRS assets for these demonstrations, CANMPER also will work with research teams to identify opportunities to perform R and D demonstrations at other facilities. Unique to this approach is the fact that these SRS assets will continue to accomplish DOE's critical

  5. The upgradation of Aditya Tokamak

    International Nuclear Information System (INIS)

    Aditya Tokamak is the first Indian tokamak, indigenously built and commissioned at the Institute for Plasma Research, Gandhinagar, Gujarat, India, in September, 1989. Aditya Tokamak has been in operation since more than 25 years. More than 30,000 discharges are taken and a large number of experiments are carried out, with plasma current ranging from 50 KA to 150 KA, lasting for 100 to 250 milliseconds. Various types of wall conditioning techniques and different hot plasma diagnostics are tested and operated on Aditya Tokamak. The experiments for turbulent particle transport and turbulence in the edge plasma, gas puffing, lithium coating, mitigation, plasma disruption, limiter and electron biasing, runaway discharges etc. led to many interesting results contributing immensely to the world of thermonuclear fusion. Experiments on Pre-ionization and Plasma heating by ICRH and ECRH are also worked out. The scientific objectives of Aditya tokamak Upgrade include Low loop voltage plasma start-up with strong pre-ionization having a good plasma control system. The upgrade is designed keeping in mind the experiments, disruption mitigation studies relevant to future fusion devices, runway mitigation studies, demonstration of Radio-frequency heating and current drive etc. This upgraded Aditya tokamak will be used for basic studies on plasma confinement and scaling to larger devices, development and testing of new diagnostics etc. This machine will be easily accessible compared to SST-1 and will be very useful for generation of technical and scientific expertise for future fusion devices. In this paper, especial features of the upgrade including various aspects of designing of new components for Aditya Upgrade tokamak is presented

  6. 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.

  7. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  8. 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

  9. Study of the generation of non-inductive current in Tore Supra and application to operational scenarios of a continuous tokamak

    International Nuclear Information System (INIS)

    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)

  10. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    International Nuclear Information System (INIS)

    Burning plasma simulation in tokamak (TR), spherical tokamak (ST) and helical (HR) reactors were carried out focusing on Internal Transport Barrier (ITB) plasma operations using the TOTAL-T (Toroidal Transport Analysis Linkage - Tokamak) code coupled with GLF23 turbulent transport code and NCLASS neoclassical transport codes, and TOTAL-H (Helical) code with multi-helicity helical ripple transport analysis code. The effectiveness of these ITB transport coefficients is checked using experimental data of JT-60U and LHD. It clarified the requirement of deep penetration of high-field-side (HFS) pellet injection fueling to realize steady-state advanced burning operation in TR and ST. The neoclassical ripple transport plays an important role on the ITB operation in HR. Moreover, economical and environmental assessments were performed for these three type reactors by the PEC (Physics Engineering and Cost) system code in the case of four blanket designs (Li/V, Flibe/FS(Ferritic Steel), LiPb/SiC, FF(Fission- Fusion) Hybrid). In the present analysis, maximum field of superconducting coil is assumed 13 T, instead of maximum normal conductor strength of 8T in ST reactor. The tolerable neutron wall fluence is assumed 20 MW.Yr/m"2 in the case of LiPb/SiC blanket system, which determines the replacement cycle of blanket modules. As for cost analysis, the fusion island (FI) cost of ST-1 is lowest. However, its fusion thermal power is largest and the TR is superior in cost of electricity (COE). Among four blanket designs Flibe/FS is superior in cost, because ferritic steel (FS) is much cheaper than vanadium (V). The life-cycle CO2 emission amount per output electric power and the energy payback ratio are also evaluated. The ST reactor is favorable in CO2 emission reduction, because rather compact and simple normal conducting coil system is adopted here. The ST and TR need more frequent blanket exchanges than HR with lower neutron wall load. However, HR is still expensive and has

  11. 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.

  12. Improved density measurement by FIR laser interferometer on EAST tokamak

    International Nuclear Information System (INIS)

    Highlights: • In 2012, the water-cooling Mo wall was installed in EAST. • A schottky barrier diode detector is designed and used on EAST for the first time. • The three-channel far-infrared laser interferometer can measure the electron density. • The improved measurement and latest experiment results are reported. • The signal we get in this experiment campaign is much better than we got in 2010. -- Abstract: A three-channel far-infrared (FIR) hydrogen cyanide (HCN) laser interferometer is in operation since 2010 to measure the line averaged electron density on experimental advanced superconducting tokamak (EAST). The HCN laser signal is improved by means of a new schottky barrier diode (SBD) detector. The improved measurement and latest experiment results of the three-channel FIR laser interferometer on EAST tokamak are reported

  13. 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

  14. 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)

  15. Metrology measurements for Aditya tokamak upgradation

    International Nuclear Information System (INIS)

    After 25 years of Aditya tokamak (midsized, air-core, R0= 75 cm, a = 25 cm) operation achieving high temperature circular plasmas in limiter configuration, upgrading it to Aditya-U tokamak with divertor configuration has been planned and the upgradation is under progress. The upgradation process include dismantling of the existing Aditya tokamak to its base level and re-erect it by placing new subsystems like new vacuum vessel of circular cross-section, new buckling cylinder etc. Apposite alignment of subsystems, mainly all the magnetic coil systems in all grades and scales of tokamak is very crucial and essential, as misaligned magnetic coil system scan generate error magnetic fields, which can significantly impact the plasma formation and sustainment in a tokamak. With this motivation, position and alignment measurement of the existing magnetic coils and structural components of ADITYA tokamak is carried out for the very first time with the optical metrology instrument. Prior to carrying out measurement exercise, machine datum has been transferred to the reference on the wall of tokamak hall using five-point laser and the machine center has been transformed to the four wall of tokamak hall. All position measurements are done with respect to machine major axis in cylindrical geometry. Measurement includes existing radial (R) and elevation (Z) positions of all magnetic coils and various structural components within the accuracy of ± 1 mm. More than 5000 data points are recorded using optical metrology instrument. Again the elevation references are transferred to the primary network established and the angular references are transformed on the floor of the tokamak hall. These results will serve as ready reference for reassembly and alignment of Aditya - Upgrade tokamak. In this paper detailed position measurements of different subsystems of old Aditya tokamak and the relocation of them along with new ones using the optical metrology instruments will be presented

  16. Operator information displays for normal operation and fault management of an advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    The paper describes the design intent, organisation and implementation of operator displays for a 1300 MWe Advanced Gas-Cooled Reactor employing a high level of automation, computer based displays and discrete devices incorporated into a structured information system. A description of the main plant and post-cooling system is given and the provisions for automatic initiation of the post-trip systems are related to claims made on operator action. A statement on control room staffing and task allocation is included in order to identify the roles of the various display systems. The structure of the Operator and Supervisor information system is described and the hierarchy of data and alarm displays is discussed. The role of discrete displays is discussed both in relation to the computer based displays and as a stand-alone set of essential information providing for safe operation in the rare event of total loss of computer based displays. The extent and organisation of the computer based displays is described with emphasis on the allocation of data to functional groups in order to aid fault diagnosis and management. The role and implementation of overview displays is described. An important part of the overall information system is the reactor post trip cooling system display. Taking the form of a functional mimic, the display provides a comprehensive summary of post trip reactivity, primary circuit integrity, reactor heat removal and cooling system status and security. The display is used to monitor the effectiveness and adequacy of the post-trip functions and provides an overview of plant abnormalities which are then investigated using the systems described above. The paper concludes with a review of areas of current and future development. (author)

  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. 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.)

  19. 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.

  20. Conceptual design of advanced steady-state tokamak reactor (A-SSTR2) - Compact and safety oriented commercial power plant

    International Nuclear Information System (INIS)

    Based on the last decade JAERI reactor design studies, the advanced commercial reactor concept (A-SSTR2) which meets both economical and environmental requirements has been proposed. The A-SSTR2 is a compact power reactor (Rp=6.2m, ap=1.5m, Ip=12MA) with a high fusion power (Pf=4GW) and a net thermal efficiency of 51%. The machine configuration is simplified by eliminating a center solenoid (CS) coil system. SiC/SiC composite for blanket structure material, helium gas cooling with pressure of 10MPa and outlet temperature of 900 deg. C, and TiH2 for bulk shield material are introduced. For the toroidal field (TF) coil, a high temperature (TC) superconducting wire made of bismuth with the maximum field of 23T and the critical current density of 1000A/mm2 at a temperature of 20K is applied. In spite of the CS-less configuration, a computer simulation gives a satisfactory plasma equilibria, plasma initiation process and current ramp up scenario. (author)

  1. Tokamak equilibria with nearly zero central current: the current hole

    International Nuclear Information System (INIS)

    The observation of stable sustainment of the 'current hole', namely the nearly zero current density region in the central part of a tokamak plasma, has opened a new class of configurations in tokamak plasmas, and a variety of research from the viewpoints of equilibrium, magnetohydrodynamics (MHD) stability, particle orbits and radial transport has been generated. Some theories and codes have been tested and extended by being applied to extreme conditions in the current hole with very weak poloidal field. The current hole is generated due to a transient negative toroidal electric field established when a large off-axis non-inductive current is rapidly formed. It has been observed in high confinement plasmas with a large fraction of bootstrap current in advanced tokamak operation. The current hole is very stiff against current drive, which suggests that it is a saturated or self-organized system. Appearance of the current hole in ITER and DEMO would be expected in some of the operation scenarios, and its influence and its control methods have been studied. Results of experimental and theoretical studies on the current hole are reviewed. (review article)

  2. 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

  3. 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

  4. Development of integrated real-time control of internal transport barriers in advanced operation scenarios on JET

    International Nuclear Information System (INIS)

    Full text: An important experimental programme is in progress on JET to investigate plasma control schemes which could enable advanced tokamak operation scenarios to eventually provide steady state burning plasmas in ITER. In particular, we have recently developed a multi-variable model-based technique for the simultaneous control of the current and pressure profiles in discharges with internal transport barriers (ITB), using lower hybrid current drive (LHCD) together with neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH). The proposed control scheme relies on the experimental identification of a linearized integral model operator and retains the intrinsic couplings between the plasma parameter profiles, as well as their distributed nature. A first set of experiments was performed in the low-density/ low-power LH-driven phase of the JET advanced scenarios, using only one actuator (LHCD) and a lumped-parameter version of the algorithm. Several requested steady state magnetic equilibria (defined by the values of the safety factor at 5 specified radii) were thus obtained and sustained for about 7s, up to full relaxation of the ohmic current throughout the plasma. Then, more interestingly in view of high power operation, a second set of experiments was dedicated to the control of the q-profile during the intense heating phase of advanced scenarios. The safety factor profile was also shown to approach a requested profile (again defined by its values at 5 radii) within about 5s. The achieved plasma equilibrium state was close to steady state. Finally, during the recent high power experimental campaign, experiments have been conducted in a 3T/1.7MA plasma, achieving for the first time the simultaneous control of the current density and electron temperature profiles in ITB plasmas. Here, the distributed-parameter version of the algorithm was used, with 3 actuators (LHCD, NBI and ICRH), and 8 output parameters [the profiles are projected upon 5 cubic

  5. Termoska pro tokamak

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2014-01-01

    Roč. 7, prosinec (2014), s. 16-17 Institutional support: RVO:61389021 Keywords : fusion * tokamak * cryostat * ITER Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1604-termoska-pro-tokamak

  6. 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.

  7. PPPL tokamak program

    International Nuclear Information System (INIS)

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  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. First observation of a new zonal-flow cycle state in the H-mode transport barrier of the experimental advanced superconducting Tokamak

    DEFF Research Database (Denmark)

    Xu, G.S.; Wang, H. Q.; Wan, B. N.; Guo, H. Y.; Naulin, Volker; Diamond, P. H.; Tynan, G. R.; Xu, M.; Yan, Ning; Zhang, W.; Chang, J. F.; Wang, L.; Chen, R.; Liu, S. C.; Ding, S. Y.; Shao, L. M.; Xiong, H.; Zhao, H. L.

    2012-01-01

    A new turbulence-flow cycle state has been discovered after the formation of a transport barrier in the H-mode plasma edge during a quiescent phase on the EAST superconducting tokamak. Zonal-flow modulation of high-frequency-broadband (0.05-1MHz) turbulence was observed in the steep-gradient region...

  10. 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.

  11. 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

  12. A Key to Improved Ion Core Confinement in the JET Tokamak: Ion Stiffness Mitigation due to Combined Plasma Rotation and Low Magnetic Shear

    DEFF Research Database (Denmark)

    Mantica, P.; Challis, C.; Peeters, A.G.; Strintzi, D.; Tala, T.; Tsalas, M.; deVries, P.C.; Baiocchi, B.; Baruzzo, M.; Bizarro, J.; Buratti, P.; Citrin, J.; Colyer, G.; Crisanti, F.; Garbet, X.; Giroud, C.; Hawkes, N.; Hobirk, J.; Imbeaux, F.; Joffrin, E.; Johnson, T.; Lerche, E.; Mailloux, J.; Naulin, Volker; Salmi, A.; Sozzi, C.; Staebler, G.; Van Eester, D.; Versloot, T.; Weiland, J.

    2011-01-01

    for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future...

  13. LOFT advanced control room operator diagnostic and display system (ODDS)

    International Nuclear Information System (INIS)

    The Loss-of-Fluid Test (LOFT) Reactor Facility in Idaho includes a highly instrumented nuclear reactor operated by the Department of Energy for the purpose of establishing nuclear safety requirements. The results of the development and installation into LOFT of an Operator Diagnostic and Display System (ODDS) are presented. The ODDS is a computer-based graphics display system centered around a PRIME 550 computer with several RAMTEK color graphic display units located within the control room and available to the reactor operators. Use of computer-based color graphics to aid the reactor operator is discussed. A detailed hardware description of the LOFT data system and the ODDS is presented. Methods and problems of backfitting the ODDS equipment into the LOFT plant are discussed

  14. Super H-mode: theoretical prediction and initial observations of a new high performance regime for tokamak operation

    Science.gov (United States)

    Snyder, P. B.; Solomon, W. M.; Burrell, K. H.; Garofalo, A. M.; Grierson, B. A.; Groebner, R. J.; Leonard, A. W.; Nazikian, R.; Osborne, T. H.; Belli, E. A.; Candy, J.; Wilson, H. R.

    2015-08-01

    A new ‘Super H-mode’ regime is predicted, which enables pedestal height and predicted fusion performance substantially higher than for H-mode operation. This new regime is predicted to exist by the EPED pedestal model, which calculates criticality constraints for peeling-ballooning and kinetic ballooning modes, and combines them to predict the pedestal height and width. EPED usually predicts a single (‘H-mode’) pedestal solution for each set of input parameters, however, in strongly shaped plasmas above a critical density, multiple pedestal solutions are found, including the standard ‘H-mode’ solution, and a ‘Super H-Mode’ solution at substantially larger pedestal height and width. The Super H-mode regime is predicted to be accessible by controlling the trajectory of the density, and to increase fusion performance for ITER, as well as for DEMO designs with strong shaping. A set of experiments on DIII-D has identified the predicted Super H-mode regime, and finds pedestal height and width, and their variation with density, in good agreement with theoretical predictions from the EPED model. The very high pedestal enables operation at high global beta and high confinement, including the highest normalized beta achieved on DIII-D with a quiescent edge.

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. 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)

  2. Analysis of neutral hydrogenic emission spectra in a tokamak

    International Nuclear Information System (INIS)

    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

  3. Tropical Cyclones: Forecasting Advances, Science Opportunities and Operational Challenges

    Science.gov (United States)

    Bosart, L. F.

    2014-12-01

    Although skill in forecasting the tracks of tropical cyclones (TCs) by operational forecast centers have improved steadily over the last 25 years, corresponding forecasts of TC intensity have shown little improvement until recently. These recent improvements in TC intensity forecasts appear to be related to a combination of better data assimilation, improved physics, and increased resolution in global operational numerical weather prediction models and new knowledge gained from a variety of recent TC-related field programs such as BGRIP, IFEX,and PREDICT. The first part of this presentation will briefly review the state of the art of TC track and intensity forecasting. The bulk of this presentation will address important TC-related science and operational challenges. These challenges include: 1) determining the physical processes that govern TC clustering, mutually interacting TCs, and the existence of different TC genesis pathways, 2) establishing how tropical-midlatitude interactions associated with recurving and transitioning (extratropical transition) TCs can trigger downstream baroclinic development, the subsequent formation of eastward-propagating Rossby wave trains, and the ensuing occurrence of extreme weather events well downstream, and 3) identifying critical TC-related forecast problems such as forecasts of the timing and extent of coastal storm surges and inland flooding associated with landfalling TCs). These important science and operational challenges will be illustrated with brief case studies.

  4. 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

  5. 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.

  6. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water...

  7. Burn Control Mechanisms in Tokamaks

    Science.gov (United States)

    Hill, Maxwell; Stacey, Weston

    2013-10-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamaks, especially those used as a neutron source for fusion-fission hybrid reactors, such as the Subcritical Advanced Burner Reactor (SABR) concept. At Georgia Tech, we are developing a new burning plasma dynamics code to investigate passive safety mechanisms that could prevent power excursions in tokamak reactors. This code solves the coupled set of balance equations governing burning plasmas in conjunction with a two-point SOL-divertor model. Predictions have been benchmarked against data from DIII-D. We are examining several potential negative feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instabilities, iii) the degradation of alpha-particle confinement resulting from ripples in the toroidal field, iv) modifications to the radial current profile, v) ``divertor choking'' and vi) Type 1 ELMs.

  8. Neutral beam injection system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B.H.; Lee, K.W.; Chung, K.S.; Oh, B.H.; Cho, Y.S.; Bae, Y.D.; Han, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    The NBI system for KSTAR (Korean Superconducting Tokamak Advanced Research) has been designed based on conventional positive ion beam technology. One beam line consists of three ion sources, three neutralizers, one bending magnet, and one drift tube. This system will deliver 8 MW deuterium beam to KSTAR plasma in normal operation to support the advanced experiments on heating, current drive and profile control. The key technical issues in this design were high power ion source(120 kV, 65 A), long pulse operation (300 seconds; world record is 30 sec), and beam rotation from vertical to horizontal direction. The suggested important R and D points on ion source and beam line components are also included. (author). 7 refs., 27 figs., 1 tab.

  9. 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.

  10. 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.)

  11. 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

  12. 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...

  13. 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)

  14. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

  15. Mechanical design and operating behaviour of advanced LWR fuel elements

    International Nuclear Information System (INIS)

    The development of fuel elements for pressurized and and boiling water reactors during the last years was marked by a reduction of the fuel cycle costs with security and reliability in operation remaining constant. The heightening of fuel discharge burnup and the improvement of neutron economy contributed essentially to that. The latter had been achieved by a reduction of the parasitic absorption within the fuel element and the leakage of neutrons of the reactor cores. These improvements could be obtained under complete observance of the safety-relevant requirements. Due to the change to fuel elements with a higher number of rods and correspondingly lower rod power it was even possible to raise the security margins partly. A survey of the state of experiences of Siemens/KWU is given. (orig./DG)

  16. 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.

  17. 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

  18. 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)

  19. 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...

  20. 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)

  1. 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

  2. Steady-state fully noninductive operation with electron cyclotron current drive and current profile control in the tokamak á configuration variable(TVC)*

    Czech Academy of Sciences Publication Activity Database

    Sauter, O.; Angioni, C.; Coda, S.; Gomez, P.; Goodman, T. P.; Henderson, M. A.; Hofmann, F.; Hogge, J. P.; Moret, J. M.; Nikkola, P.; Pietrzyk, Z. A.; Weisen, H.; Alberti, S.; Appert, K.; Bakos, J.; Behn, R.; Blanchard, P.; Bosshard, P.; Chavan, R.; Condrea, I.; Degeling, A.; Duval, B. P.; Fasel, D.; Favez, J. Y.; Favre, A.; Furno, I.; Kayruthdinov, R. R.; Lavanchy, P.; Lister, J. B.; Llobet, X.; Lukash, V. E.; Gorgerat, P.; Isoz, P. F.; Joye, B.; Magnin, J. C.; Manini, A.; Marlétaz, B.; Marmillod, P.; Martin, Y. R.; Martynov, An.; Mayor, J. M.; Minardi, E.; Mlynar, J.; Paris, P. J.; Perez, A.; Peysson, Y. R.; Piffl, Vojtěch; Pitts, R. A.; Pochelon, A.; Reimerdes, H.; Rommers, J. H.; Scavino, E.; Sushkov, A.; Tonetti, G.; Tran, M. Q.; Zabolotsky, A.

    2001-01-01

    Roč. 8, č. 5 (2001), s. 2199-2207. ISSN 1070-664X. [Annual Meeting of the APS Division of Plasma Physics with the 10th International Congress on Plasma Physics/42nd./. Quebec City, Quebec, 23.10.2000-27.10.2000] Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.223, year: 2001

  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. Automated Fault Detection for DIII-D Tokamak Experiments

    International Nuclear Information System (INIS)

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

  5. Recent experiments and upgradation plans for Aditya Tokamak

    International Nuclear Information System (INIS)

    Several experiments relevant to the operation of future big tokamaks including ITER and also contributing significantly to the tokamak based thermonuclear fusion research have been carried out in Aditya tokamak recently. Low loop voltage start-up of plasma current has been successfully obtained with ICR and ECR preionization. Reduced runaway generation is achieved by applying a local vertical magnetic field at one toroidal location. Plasma disruptions, a sudden loss of equilibrium and confinement, has been successfully mitigated through application of bias voltage on a Molybdenum (Mo) electrode placed inside the last closed flux surface (LCFS). Extensive studies on plasma flows, effect of gas puff on flows in the Scrape off layer (SOL) and impurity transport has been carried out. Effect of Helium glow discharge cleaning (GDC) on partial pressures and plasma parameters have also been studied for plasma performance improvement. To contribute more to the bigger tokamaks operated in the divertor configuration, the existing Aditya tokamak with limiter configuration, which is in operation for 24 years, is planned to be upgraded to a divertor machine. The main aim of the Aditya-U tokamak is to have a small/mid-size tokamak with divertor operation and higher duty cycle, which will be test bed for new diagnostics, the students can be trained and those experiments can be tried out which are not desirable in big tokamaks, such as runaway mitigation and disruption control. Details of experimental results and upgradation plan will be discussed in the talk. (author)

  6. 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)

  7. 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.

  8. ACR-1000: Advanced I and C and IT systems to enhance operations and maintenance

    International Nuclear Information System (INIS)

    The Advanced CANDU Reactor ACR-1000 is a 1200-MWe-class Generation III+ nuclear power plant designed by Atomic Energy of Canada Limited (AECL). Its design is evolutionary, starting with a strong base of proven CANDU reactor technology coupled with thoroughly demonstrated innovative features to enhance economics, safety, operability and maintainability. Two key design strategies were to expand the Instrumentation and Control (I and C) and Information Technology (IT) systems, and improve Operations and Maintenance (O and M) capability. AECL has developed ACR-1000 I and C and IT systems, including SMART CANDU, to improve the timeliness, the quality and the integration of the information made available to plant operators and engineers. These systems use automatic data mining and present organized and analyzed data to operators and engineers to facilitate diagnostics and reduce mental burden - thus increasing the ability to make proactive and informed decisions affecting plant operation. Major advances have also been made in designing and ACR-1000 control room itself. Additionally, direct feedback from CANDU plant operators on enhancing the ACR-1000's operability has permitted optimization of on-line maintenance and facilitated and reduced off-line maintenance. New features have been designed into the plant to reduce operating risk and reduce costs. This paper outlines how IT and O and M advances have enabled the ACR-1000 to meet and exceed performance and operability targets. (author)

  9. 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

  10. 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.

  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. Tokamak Physics Experiment divertor design

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m2. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services

  13. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-βN tokamak reactors. (author)

  14. 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

  15. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

  16. Three novel tokamak plasma regimes in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Furth, H.P.

    1985-10-01

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region.

  17. A Modern Advanced Hill Cipher Involving a Permuted Key and Modular Arithmetic Addition Operation

    Directory of Open Access Journals (Sweden)

    V.U.K. Sastry

    2011-05-01

    Full Text Available In this paper we have devoted our attention to the study of a block cipher by generalizing advanced Hill cipher by including a permuted key. In this analysis we find that the iteration process, the mix operation and the modular arithmetic operation involved in the cipher mixes the binary bits of the key and the plaintext in a thorough manner. The avalanche effect and the cryptanalysis markedly indicate that the cipher is a strong one.

  18. 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.

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

    International Nuclear Information System (INIS)

    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 μs. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  6. 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.

  7. 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.

  8. 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)

  9. Using CONFIG for Simulation of Operation of Water Recovery Subsystems for Advanced Control Software Evaluation

    Science.gov (United States)

    Malin, Jane T.; Flores, Luis; Fleming, Land; Throop, Daiv

    2002-01-01

    A hybrid discrete/continuous simulation tool, CONFIG, has been developed to support evaluation of the operability life support systems. CON FIG simulates operations scenarios in which flows and pressures change continuously while system reconfigurations occur as discrete events. In simulations, intelligent control software can interact dynamically with hardware system models. CONFIG simulations have been used to evaluate control software and intelligent agents for automating life support systems operations. A CON FIG model of an advanced biological water recovery system has been developed to interact with intelligent control software that is being used in a water system test at NASA Johnson Space Center

  10. 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.

  11. Advanced Launch System (ALS) actuation and power systems impact operability and cost

    Science.gov (United States)

    Sundberg, Gale R.

    1990-01-01

    To obtain the Advanced Launch System (ALS) primary goals of reduced costs and improved operability, there must be significant reductions in the launch operations and servicing requirements relative to current vehicle designs and practices. One of the primary methods for achieving these goals is by using vehicle electrical power system and controls for all actuation and avionics requirements. A brief status review of the ALS and its associated Advanced Development Program is presented to demonstrate maturation of those technologies that will help meet the overall operability and cost goals. The electric power and actuation systems are highlighted as a specific technology ready not only to meet the stringent ALS goals (cryogenic field valves and thrust vector controls with peak power demands to 75 hp), but also those of other launch vehicles, military and civilian aircraft, lunar/Martian vehicles, and a multitude of commercial applications.

  12. Advanced launch system (ALS) - Electrical actuation and power systems improve operability and cost picture

    Science.gov (United States)

    Sundberg, Gale R.

    1990-01-01

    To obtain the Advanced Launch System (ALS) primary goals of reduced costs and improved operability, there must be significant reductions in the launch operations and servicing requirements relative to current vehicle designs and practices. One of the primary methods for achieving these goals is by using vehicle electrrical power system and controls for all aviation and avionics requirements. A brief status review of the ALS and its associated Advanced Development Program is presented to demonstrate maturation of those technologies that will help meet the overall operability and cost goals. The electric power and actuation systems are highlighted as a sdpecific technology ready not only to meet the stringent ALS goals (cryogenic field valves and thrust vector controls with peak power demands to 75 hp), but also those of other launch vehicles, military ans civilian aircraft, lunar/Martian vehicles, and a multitude of comercial applications.

  13. Advanced Launch System (ALS): Electrical actuation and power systems improve operability and cost picture

    Science.gov (United States)

    Sundberg, Gale R.

    1990-01-01

    To obtain the Advanced Launch System (ALS) primary goals of reduced costs and improved operability, there must be significant reductions in the launch operations and servicing requirements relative to current vehicle designs and practices. One of the primary methods for achieving these goals is by using vehicle electrical power system and controls for all actuation and avionics requirements. A brief status review of the ALS and its associated Advanced Development Program is presented to demonstrate maturation of those technologies that will help meet the overall operability and cost goals. The electric power and actuation systems are highlighted as a specific technology ready not only to meet the stringent ALS goals (cryogenic field valves and thrust vector controls with peak power demands to 75 hp), but also those of other launch vehicles, military and civilian aircraft, lunar/Martian vehicles, and a multitude of commercial applications.

  14. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

  15. 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

  16. Small tokamaks for fusion technology testing

    International Nuclear Information System (INIS)

    Small steady-state tokamaks for testing divertors and fusion nuclear technologies are considered. Based on present physics and technology data and explanation to reduce R0/a, H-D-fueled tokamaks with R0 ∼ 0.6--0.75 m, R0/a ∼ 1.8--2.5, and Bt0 ∼ 1.4--2.2 T can be driven with Ptot ∼ 4.5 MW to maintain Ip ∼ 0.5 MA and produce the ITER-level plasma edge and divertor conditions. Given an adequate steady-state divertor solution and Q∼1 operation based on fusion through the suprathermal component, D-T-fueled tokamaks with R0 ∼ 0.8 m, R0/a ∼ 2, and Bt0 ∼ 4 T can be driven with Ptot ∼ 15 MW to maintain Ip ∼ 4.6 MA and produce an peak neutron wall load WL ∼ 1 MW/m2. Such devices appear possible if the plasma properties at the power R0/a remain tokamak-like and, for the D-T case, can unshielded center core is feasible. The use of a single conductor as the inboard leg of the toroidal field coils for this purpose is discussed. The physics issues and the design features are identified for such tokamaks with a testing duty for factor goal of 10--20%

  17. The Thor tokamak experiment

    International Nuclear Information System (INIS)

    The main characteristics of the plasma produced in Thor tokamak discharges are described. The machine performances are outlined and the experimental results relevant to the equilibrium, the stability and the control of the discharge regimes are discussed in detail. (author)

  18. Modular tokamak magnetic system

    Science.gov (United States)

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  19. 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

  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. Experiences on vacuum conditioning in the cryostat of KSTAR tokamak

    International Nuclear Information System (INIS)

    Highlights: ► The vacuum of the cryostat has been stably maintained during the KSTAR operation. ► The detected cold leak at the PF/CS coils and CS structure. ► The present helium leak makes no issue for the cryostat operation. -- Abstract: Korea Superconducting Tokamak Advanced Research (KSTAR) device has been successfully operated for the plasma experiment from KSTAR 1st campaign to 4th campaign. The main pumping system for the cryostat has to maintain the target pressure below 1.0 × 10−4 mbar at room temperature and 1.0 × 10−5 mbar at extremely low temperature for the plasma experiment against the air leak coming from ports of vessel and/or the helium leak from cooling loops in the cryostat. No leak has been detected at room temperature. Unexpectedly, the cold-leak appeared in the cryostat at temperature around 50 K during the cool-down in the KSTAR 2nd campaign. We carefully analyzed the characteristics of detected cold leak because it can cause the increase of the base pressure in the cryostat. After the cool-down, the leak detection was performed to locate the position and size of the leak by the pressurizing the loops. As a result, it is found that the cold leak was located at cooling loops for PF/CS coils and CS structure. Nevertheless, the vacuum inside the cryostat was well maintained below 6.0 × 10−8 mbar during the entire operation period. The impact of the He-leak in present status on the plasma operation is negligible. However, we have found that the leak rate increases as a function of time. Therefore careful monitoring on cold-leak is an important technical issue for the operation of superconducting tokamak

  2. 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

  3. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs.

  4. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  5. 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

  6. 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)

  7. 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)

  8. 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

  9. Predictive Modeling of Tokamak Configurations*

    Science.gov (United States)

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

  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. Romanian research in the field of Tokamak fusion reactors

    International Nuclear Information System (INIS)

    To re-create the conditions of the sun and stars for the production of fusion energy on earth, scientists most accomplish three major tasks. They have already passed the first task by achieving the necessary temperatures. In same cases, they have attained temperatures as high as 510 million degrees, 20 times more then the temperature at the center of the sun. Secondly, they need to demonstrate sustained reactions where substantial amounts of energy are produced. The third major milestone for fusion would be operation of a demonstration fusion power plant. Many different magnetic confined schemes have been studied. The one which is receiving the greatest attention in the international magnetic fusion energy programme is the tokamak concept, and represents actually the most advanced fusion devices. The advantage of fusion are: - abundant fuel supply; - no risk of a nuclear accident; - no air pollution; - no high-level nuclear waste; - no generation of weapons material. The present objectives and research priorities of the fusion community are: - continuation of ongoing research; - concept improvements; - long term technology. Our research programme in the field of tokamak fusion reactions is performed mainly in the frame of international cooperation with 'I.V. Kurchatov' Nuclear Fusion Institute from Moscow, Institute of Applied Mathematics from Grenoble, Research Center from Cadarache, 'Max-Planck' Institute for Plasma Physics from Garching at Munich and Columbia University from New York. The activities carried out under our programme are closely coordinated with those of the European Atomic Energy Community and are related to current problems concerning equilibrium, stability, transport and diagnostics of tokamak plasmas. Our results are mentioned in the International Atomic Energy Agency's World Survey of Activities in Controlled Fusion Research in 1997 and the European Community's Reports EUR FUR BRU from 1993 and 1996. (author)

  12. 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

  13. High βp bootstrap tokamak reactor

    International Nuclear Information System (INIS)

    Basic characteristics of a steady state tokamak fusion reactor is presented. The minimum required energy multiplication factor Q is found to be 20 to 30 for the feasibility of the fusion reactor. Such a high Q steady state tokamak operation is possible, within our present knowledge of the operational constraints and the current drive physics, when a large fraction of the plasma current is carried by the bootstrap current. Operation at high βp (≥2.0) and high qψ (=4-5) with relatively small εβp (3) and fusion output power (2.5 GW) and is consistent with the present knowledges of the plasma physics of the tokamak, namely the Troyon limit, the energy confinement scalings, the bootstrap current, the current drive efficiency (NB current drive with the total power of 70 MW and the beam energy of 1 MeV) with a favorable aspect on the formation of the cold and dense diverter plasma-condition. From the economical aspect of the tokamak fusion reactor, a more compact reactor is favorable. The use of the high field magnet with Bmax = 16T (for example Ti-doped Nb3Sn conductor) enables to reduce the total machine size to 50% of the above-described conventional design, namely Rp = 7m, Vp = 760m-3, PF = 2.8 GW. (author)

  14. 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.

  15. 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)

  16. 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

  17. A Dynamical-System Description of ITB Oscillations in Tokamaks

    International Nuclear Information System (INIS)

    Internal transport barriers (ITBs) are a key ingredient in advanced tokamak scenarios aiming at steady-state operation and, as presented in the last FPPT (Kathmandu), they sometimes exhibit an oscillatory behavior. In this communication, the interplay between the temperature and the non-inductive current, together with its role on ITB dynamics in tokamak plasmas, is studied. A set of two ordinary differential equations is derived from the reduction of the fluid-like radial energy and current diffusion equations, in which a mixed Bohm-gyroBohm energy diffusivity, dependent on the local magnetic shear, is used. The bifurcation analysis of the resulting dynamical system is attempted. Depending on the values of a set of relevant parameters, the system’s orbits are shown to possess one of three distinct limit sets: two fixed points, one characteristic of a lowconfinement state and the other characteristic of a high-confinement state, and a limit cycle. These correspond, respectively, to the following experimentally observed states: absence of an ITB, ITB persistence, and ITB repetitive rise and fall in a sequence of oscillations. As a result, threshold conditions for the different regimes are atempted. (author)

  18. High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facility

    International Nuclear Information System (INIS)

    This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy

  19. 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.

  20. HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

    International Nuclear Information System (INIS)

    HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

  1. 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...

  2. STAR-GENERIS: A software concept for advanced information presentation of future operator aids

    International Nuclear Information System (INIS)

    Man-machine communication in power plants is increasingly relying on improved information quality provided by computerized operator support systems. It could be demonstrated that the different plant functions can be represented by one advanced software programme package, the STAR-GENERIS. The specific applications now under development include: symptom-oriented display of disturbed plant situation, post-trip analysis, integrated disturbance analysis, alarm reduction, and status surveillance of components and plant systems. (author). 2 figs

  3. Development of advanced human-machine system for plant operation and maintenance

    International Nuclear Information System (INIS)

    With the worldwide deregulation of the power industry, and the aging of the nuclear power plants (NPPs), concerns are growing over the reliability and safety of the NPPs, because the regulation of man power may lower the current high level of reliability and safety. In this paper, a concept of overall integrated plant management mechanism is proposed, in order to meet the requirements of cutting costs of NPPs and the requirements of maintaining or increasing safety and reliability. The concept is called as satellite operation maintenance center (SOMC). SOMC integrates the operation and maintenance activities of several NPP units by utilizing advanced information technologies to support cooperation activities between workers allocated at SOMC and the field workers. As for the operation activities, a framework called as Advanced Operation System (AOS) is proposed in this paper. AOS consists of three support sub-systems: dynamic operation permission system(DyOPS), supervisor information presentation system using interface agent, and crew performance evaluation system. As for the maintenance activities, a framework called as Ubiquitous-Computing-based Maintenance support System (UCMS) is proposed next. Two case studies are described, in order to show the way of how UCMS support field workers to do maintenance tasks efficiently, safely, and infallibly as well. Finally, a prospect of SOMC is shown in order to explain the way of how the technology elements developed in this project could be integrated as a whole one system to support maintenance activities of NPPs in the future. (author)

  4. 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

  5. Computer visualization for enhanced operator performance for advanced nuclear power plants

    International Nuclear Information System (INIS)

    The operators of nuclear power plants are presented with an often uncoordinated and arbitrary array of displays and controls. Information is presented in different formats and on physically dissimilar instruments. In an accident situation, an operator must be very alert to quickly diagnose and respond to the state of the plant as represented by the control room displays. Improvements in display technology and increased automation have helped reduce operator burden; however, too much automation may lead to operator apathy and decreased efficiency. A proposed approach to the human-system interface uses modern graphics technology and advances in computational power to provide a visualization or ''virtual reality'' framework for the operator. This virtual reality comprises a simulated perception of another existence, complete with three-dimensional structures, backgrounds, and objects. By placing the operator in an environment that presents an integrated, graphical, and dynamic view of the plant, his attention is directly engaged. Through computer simulation, the operator can view plant equipment, read local displays, and manipulate controls as if he were in the local area. This process not only keeps an operator involved in plant operation and testing procedures, but also reduces personnel exposure. In addition, operator stress is reduced because, with realistic views of plant areas and equipment, the status of the plant can be accurately grasped without interpreting a large number of displays. Since a single operator can quickly ''visit'' many different plant areas without physically moving from the control room, these techniques are useful in reducing labor requirements for surveillance and maintenance activities. This concept requires a plant dynamic model continuously updated via real-time process monitoring. This model interacts with a three-dimensional, solid-model architectural configuration of the physical plant

  6. 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.

  7. 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.

  8. 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

  9. Advancing data assimilation in operational hydrologic forecasting: progresses, challenges, and emerging opportunities

    Directory of Open Access Journals (Sweden)

    Y. Liu

    2012-10-01

    Full Text Available 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

  10. Advancing data assimilation in operational hydrologic forecasting: progresses, challenges, and emerging opportunities

    Directory of Open Access Journals (Sweden)

    Y. Liu

    2012-03-01

    Full Text Available 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 into operational forecast systems to improve the skill of forecasts to better inform 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 considerations 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 modelling and DA tools or frameworks, and through fostering collaborative efforts

  11. Material erosion and migration in tokamaks

    International Nuclear Information System (INIS)

    The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the subsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasma-facing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined

  12. 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

  13. Integration of LHCD and IBW heating for high-performance discharges under steady-state operation in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Significant progress in obtaining high-performance discharges under quasi-steady-states in the HT-7 superconducting tokamak has been realized since the last IAEA meeting. In relation to the previous experiments, various features of the non-inductive current driven, heating, profile control, MHD stabilization and edge physics are integrated and optimized to achieve steady-state high-performance discharges. Both on-axis and off-axis electron heating with global peaked and locally steepened electron pressure profiles were realized with improved confinement if the ion Bernstein wave (IBW) resonant layer was properly selected. Stabilization of MHD instabilities was demonstrated by off-axis IBW heating. The internal transport barrier structure was formed by off-axis lower hybrid current drive (LHCD). Long-pulse discharges with Te ∼ 1 keV and central density ∼1 x 1019 m-3 were obtained with a duration of 10-20 s. A combination of IBW heating and LHCD produced a broadened current density profile, which may be a signature of the synergy effect between two waves. Experimental results show that features of IBW in controlling electron pressure profile can be integrated into LHCD target plasmas. HT-7 has produced a variety of discharges with βN x H89 > 1-4 for durations of several to several tens of energy confinement times with a non-inductive driven current of 50-80% by optimizing the IBW heating and LHCD and avoiding MHD activities. (author)

  14. Efficiency improvement of nuclear power plant operation: the significant role of advanced nuclear fuel technologies

    International Nuclear Information System (INIS)

    Due to the increased liberalisation of the power markets, nuclear power generation is being exposed to high cost reduction pressure. In this paper we highlight the role of advanced nuclear fuel technologies to reduce the fuel cycle costs and therefore increase the efficiency of nuclear power plant operation. The key factor is a more efficient utilisation of the fuel and present developments at Siemens are consequently directed at (i) further increase of batch average burnup, (ii) improvement of fuel reliability, (iii) enlargement of fuel operation margins and (iv) improvement of methods for fuel design and core analysis. As a result, the nuclear fuel cycle costs for a typical LWR have been reduced during the past decades by about US$ 35 million per year. The estimated impact of further burnup increases on the fuel cycle costs is expected to be an additional saving of US$10 - 15 million per year. Due to the fact that the fuel will operate closer to design limits, a careful approach is required when introducing advanced fuel features in reload quantities. Trust and co-operation between the fuel vendors and the utilities is a prerequisite for the common success. (authors)

  15. 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

  16. 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

  17. 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)

  18. 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.

  19. Advanced qualification process of ANAV NPP integral dynamic models for supporting plant operation and control

    International Nuclear Information System (INIS)

    Extensive technical literature exists aimed at establishing the requirements needed to qualify a Nuclear Power Plant model. Most of this literature is focused on qualifying a model for licensing uses. Less documentation is available nowadays on the requirements needed when an Integral Plant Model is used for supporting plant operation and control of an actual commercial facility, while fulfilling its goals of safety and competitiveness. For the last 15 years the Technical University of Catalonia (UPC) has been working in this field along with Asociacion Nuclear Asco-Vandellos (ANAV), which is a utility that presently runs three operating PWRs. The paper develops an advanced qualification process (AQP) of plant models for operation support, introduces the concept of plant configuration and explains how this activity complements other usual validation tasks

  20. 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

  1. Visualization and Analysis of Remote Operation involved in Advanced Conditioning Process

    International Nuclear Information System (INIS)

    The remote operation of the Advanced Spent Fuel Conditioning Process (ACP) is analyzed by using the 3D graphic simulation tools. The ACP equipment operates in intense radiation fields as well as in a high temperature. Thus, the equipment should be designed in consideration of the remote handling and maintenance. As well as suitable remote handling and maintenance technology needs to be developed along with the design of the process concepts. To develop such remote operation technology, we developed the graphic simulator which provides the capability of verifying the remote operability of the ACP without fabrication of the process equipment. In order words, by applying virtual reality to the remote maintenance operation, a remote operation task can be simulated in the graphic simulator, not in the real environment. The graphic simulator will substantially reduce the cost of the development of the remote handling and maintenance procedure as well as the process equipment, while at the same time development a remote maintenance concept that is more reliable, easier to implement, and easier to understand

  2. Objectives and design of the JT-60 superconducting tokamak

    International Nuclear Information System (INIS)

    A fully superconducting tokamak named as JT-60SC is designed for the modification program of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-beta steady-state operation in the use of low radio-activation ferritic steel in low ν and ρ regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented. (author)

  3. Development of CAPE-OPEN unit operations for advanced power systems modeling

    Energy Technology Data Exchange (ETDEWEB)

    Swensen, D.; Zitney, S.; Bockelie, M.

    2007-01-01

    Reaction Engineering International (REI) is developing a suite of CORBA CAPE-OPEN compliant unit operations to support the Department of Energy NETL Advanced Process Engineering Co-Simulation (APECS) framework for advanced power systems. One of these components will provide an interface to the GateCycle software, which is widely used in the power generation industry for process modeling. Additional CAPE-OPEN components will provide modeling capabilities of varying fidelity for equipment such as a gasifier, air separation unit, syngas cooler, solid oxide fuel cell and water-gas shift reactor. The models used in the components will be a combination of those developed by REI and proprietary models developed by members of the Project's industrial interest group.

  4. Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

    International Nuclear Information System (INIS)

    Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m2 and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust

  5. Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, L., E-mail: lwang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Dalian University of Technology, Dalian 116024 (China); Guo, H.Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); General Atomics, P. O. Box 85608, San Diego, CA 92186 (United States); Li, J.; Wan, B.N.; Gong, X.Z.; Zhang, X.D.; Hu, J.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Association EURATOM-FZJ, D-52425 Jülich (Germany); Xu, G.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zou, X.L. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Loarte, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Maingi, R.; Menard, J.E. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Luo, G.N.; Gao, X.; Hu, L.Q.; Gan, K.F.; Liu, S.C.; Wang, H.Q.; Chen, R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

    2015-08-15

    Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m{sup 2} and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust.

  6. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime

  7. 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)

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. STARPAHC - Operational findings. [Space Technology Applied to Rural Papago Advanced Health Care

    Science.gov (United States)

    Belasco, N.; Pool, S. L.

    1976-01-01

    Delivery of quality health care to passengers of extended-mission spacecraft and to remote populations on earth (a major national problem) requires extending the knowledge and skills of the physician many kilometers distant from his physical location. The STARPAHC telemedicine system accomplishes this by using physician's assistants complemented with space technology in communications, data handling, and systems engineering. It is presently in operation and undergoing a 2-year evaluation on the Papago Indian Reservation, Arizona. Results have established its feasibility as a solution for remote area health care on earth, while providing information useful to the planners of advanced manned spacecraft missions.

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. IFS Numerical Laboratory Tokamak

    International Nuclear Information System (INIS)

    A numerical laboratory of a tokamak plasma is being developed. This consists of the backbone (the overall manager in terms of the MPPL programming language), and the modularized components that can be plugged in or out for a particular run and their hierarchical arrangement. The components include various metrics for overall geometry various dynamics, field calculations, and diagnoses. 2 refs

  19. 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.

  20. Transport in gyrokinetic tokamaks

    International Nuclear Information System (INIS)

    A comprehensive study of transport in full-volume gyrokinetic (gk) simulations of ion temperature gradient driven turbulence in core tokamak plasmas is presented. Though this ''gyrokinetic tokamak'' is much simpler than experimental tokamaks, such simplicity is an asset, because a dependable nonlinear transport theory for such systems should be more attainable. Toward this end, we pursue two related lines of inquiry. (1) We study the scalings of gk tokamaks with respect to important system parameters. In contrast to real machines, the scalings of larger gk systems (a/ρs approx-gt 64) with minor radius, with current, and with a/ρs are roughly consistent with the approximate theoretical expectations for electrostatic turbulent transport which exist as yet. Smaller systems manifest quite different scalings, which aids in interpreting differing mass-scaling results in other work. (2) With the goal of developing a first-principles theory of gk transport, we use the gk data to infer the underlying transport physics. The data indicate that, of the many modes k present in the simulation, only a modest number (Nk ∼ 10) of k dominate the transport, and for each, only a handful (Np ∼ 5) of couplings to other modes p appear to be significant, implying that the essential transport physics may be described by a far simpler system than would have been expected on the basis of earlier nonlinear theory alone. Part of this analysis is the inference of the coupling coefficients Mkpq governing the nonlinear mode interactions, whose measurement from tokamak simulation data is presented here for the first time