Steady State Advanced Tokamak (SSAT): The mission and the machine
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
Plasma control issues for an advanced steady state tokamak reactor
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)
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
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.
Overview of JT-60U progress towards steady-state advanced tokamak
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)
A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution
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...
Magnetic sensor for steady state tokamak
Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1996-06-01
A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).
Real-time control of the q-profile in JET for steady state advanced tokamak operation
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)
A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution
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.
Towards steady-state tokamak operation with double transport barriers
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)
Steady State versus Pulsed Tokamak DEMO
Full text: The present report deals with a Review of problems for a Steady state(SS) DEMO, related argument is treated about the models and the present status of comparison between the characteristics of DEMO pulsed versus a Steady state device.The studied SS DEMO Models (SLIM CS, PPCS model C EU-DEMO, ARIES-RS) are analyzed from the point of view of the similarity scaling laws and critical issues for a steady state DEMO. A comparison between steady state and pulsed DEMO is therefore carried out: in this context a new set of parameters for a pulsed (6 — 8 hours pulse) DEMO is determined working below the density limit, peak temperature of 20 keV, and requiring a modest improvement in the confinement factor(HIPBy2 = 1.1) with respect to the H-mode. Both parameters density and confinement parameter are lower than the DEMO models presently considered. The concept of partially non-inductive pulsed DEMO is introduced since a pulsed DEMO needs heating and current drive tools for plasma stability and burn control. The change of the main parameter design for a DEMO working at high plasma peak temperatures Te ∼ 35 keV is analyzed: in this range the reactivity increases linearly with temperature, and a device with smaller major radius (R = 7.5 m) is compatible with high temperature. Increasing temperature is beneficial for current drive efficiency and heat load on divertor, being the synchrotron radiation one of the relevant components of the plasma emission at high temperatures and current drive efficiency increases with temperature. Technology and engineering problems are examined including efficiency and availability R&D issues for a high temperature DEMO. Fatigue and creep-fatigue effects of pulsed operations on pulsed DEMO components are considered in outline to define the R&D needed for DEMO development. (author)
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)
A Steady State Tokamak Operation by Use of Magnetic Monopoles
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...
Optimization of steady-state beam-driven tokamak reactors
Recent developments in neutral beam technology prompt us to reconsider the prospects for steady-state tokamak reactors. A mathematical reactor model is developed that includes the physics of beam-driven currents and reactor power balance, as well as reactor and beam system costs. This model is used to find the plasma temperatures that minimize the reactor cost per unit of net electrical output. The optimum plasma temperatures are nearly independent of β and are roughly twice as high as the optimum temperatures for ignited reactors. If beams of neutral deuterium atoms with near-optimum energies of 1 to 2 MeV are used to drive the current in a reactor the size of the International Tokamak Reactor, then the optimum temperatures are typically T /SUB e/ approx. = 12 to 15 keV and T /SUB i/ approx. = 17 to 21 keV for a wide range of model parameters. Net electrical output rises rapidly with increasing deuterium beam energy for E /SUB b/ less than or equal to 400 keV, but rises only slowly above E /SUB b/ about 1 MeV. We estimate that beam-driven steady-state reactors could be economically competitive with pulsed-ignition reactors if cyclic-loading problems limit the toroidal magnetic field strength of pulsed reactors to less than or equal to 85% of that allowed in steady-state reactors
Small steady-state tokamak (TST) for divertor testing
The TST is a small steady-state tokamak designed for testing diverters under conditions similar to those anticipated in future large tokamaks. An initial design has R0/a = 2.5, R0 = 0.75 m, a = 0.3 m, and Bt0 = 2.2 T with full inductive capability. With heating and current drive power of 4.5 MW, the heat flux at the plasma edge Q perpendicular can be as high as 0.3 MW/m2. Plasma currents Ip above 500 kA can be maintained by 1 MW of lower hybrid power (2.45 GHz) for average densities ne up to 3 x 1019 m-3. Additional power via ICRF (2 MW) and neutral beams (1.5 MW) maintain current for ne up to 5 x 1019 m-3. Fully demountable, actively cooled, steady-state toroidal field coils permit ample access for the auxiliary systems and diverter cassettes. The toroidal field magnets require a steady-state supply of less than 40 MW. The size and cost of the TST can be reduced by eliminating the solenoid, reducing Bt0 to 1.4 T, and lowering R0/a to 1.7. This option permits low-R0/a experimentation while maintaining the capability for testing divertors but requires successful noninductive current initiation and maintenance in the low-R0/a regime
Optimization of steady-state beam-driven tokamak reactors
Recent developments in neutral beam technology prompt us to reconsider the prospects for steady-state tokamak reactors. A mathematical reactor model is developed which includes the physics of beam-driven currents and reactor power balance, as well as reactor and beam system costs. This model is used to find the plasma temperatures which minimize the reactor cost per unit of net electrical output. The optimum plasma temperatures are nearly independent of β and are roughly twice as high as the optimum temperatures for ignited reactors. If beams of neutral deuterium atoms with near-optimum energies of 1 to 2 MeV are used to drive the current in an INTOR-sized reactor, then the optimum temperatures are typically T/sub e/ approx. = 12 to 15 keV and T/sub i/ approx. = keV for a wide range of model parameters. Net electrical output rises rapidly with increasing deuterium beam energy for E/sub b/ less than or equal to 400 keV, but rises only slowly above E/sub b/ approx. 1 MeV. We estimate that beam-driven steady-state reactors could be economically competitive with pulsed-ignition reactors if cyclic-loading problems limit the toroidal magnetic field strength of pulsed reactors to less than or equal to 85% of that allowed in steady-state reactors
Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1
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)
Concept study of the Steady State Tokamak Reactor (SSTR)
The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)
Small steady-state tokamak (TST) for divertor testing
Peng, Y.M.; Colchin, R.J.; Swain, D.W.; Nelson, B.E.; Monday, J.F. (Oak Ridge National Lab., TN (United States)); Blevins, J.; Delisle, M.; Stringer, J. (Canadian Fusion Fuels Technology Project, Mississauga, ON (Canada)); Bonoli, P.; Luckhardt, S. (Massachusetts Inst. of Tech., Cambridge, MA (United States)); Pauletti, R. (Sao Paulo Univ., SP (Brazil))
1992-01-01
The TST is a small steady-state tokamak designed for testing diverters under conditions similar to those anticipated in future large tokamaks. An initial design has R{sub 0}/a = 2.5, R{sub 0} = 0.75 m, a = 0.3 m, and Bt{sub 0} = 2.2 T with full inductive capability. With heating and current drive power of 4.5 MW, the heat flux at the plasma edge Q{perpendicular} can be as high as 0.3 MW/m{sup 2}. Plasma currents I{sub p} above 500 kA can be maintained by 1 MW of lower hybrid power (2.45 GHz) for average densities n{sub e} up to 3 {times} 10{sup 19} m{sup {minus}3}. Additional power via ICRF (2 MW) and neutral beams (1.5 MW) maintain current for n{sub e} up to 5 {times} 10{sup 19} m{sup {minus}3}. Fully demountable, actively cooled, steady-state toroidal field coils permit ample access for the auxiliary systems and diverter cassettes. The toroidal field magnets require a steady-state supply of less than 40 MW. The size and cost of the TST can be reduced by eliminating the solenoid, reducing Bt{sub 0} to 1.4 T, and lowering R{sub 0}/a to 1.7. This option permits low-R{sub 0}/a experimentation while maintaining the capability for testing divertors but requires successful noninductive current initiation and maintenance in the low-R{sub 0}/a regime.
Preliminary design study of a steady state tokamak device
Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)
Operating tokamaks with steady-state toroidal current
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
The Steady-state Tokamak (STE) Experiment is a proposed superconducting-coil, hydrogen-plasma tokamak device intended to address the integrated non-nuclear issues of steady state, high-power tokamak physics and technology. Such a facility has been called for in the US program plan for the mid 1990's, and will play a unique role in the world-wide fusion effort. Information from STE on steady-state current drive, plasma control, and high power technology will contribute significantly to the operating capabilities of future steady-state devices. This paper reviews preliminary designs and expected technological contributions to the US and world fusion reactor research from each of the above mentioned reactor systems. This document is intended as a proposal and feasibility discussion and does not include exhaustive technical reviews. 12 figs., 3 tabs
Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime
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
Anisotropic plasma with flows in tokamak: Steady state and stability
An adequate description of equilibrium and stability of anisotropic plasma with macroscopic flows in tokamaks is presented. The Chew-Goldberger-Low (CGL) approximation is consistently used to analyze anisotropic plasma dynamics. The admissible structure of a stationary flow is found to be the same as in the ideal magnetohydrodynamics with isotropic pressure (MHD), which means an allowance for the same relabeling symmetry as in ideal MHD systems with toroidally nested magnetic surfaces. A generalization of the Grad-Shafranov equation for the case of anisotropic plasma with flows confined in the axisymmetric magnetic field is derived. A variational principle was obtained, which allows for a stability analysis of anisotropic pressure plasma with flows, and takes into account the conservation laws resulting from the relabeling symmetry. This principle covers the previous stability criteria for static CGL plasma and for ideal MHD flows in isotropic plasma as well. copyright 1996 American Institute of Physics
Steady state operation of tokamaks. Proceedings of a technical committee meeting
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
Implications of rf current drive theory for next step steady-state tokamak design
Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor
Steady state tokamak equilibrium with specified magnetic axis and two magnetic null points
An analysis method of tokamak plasma equilibrium by a relaxation method with specified magnetic axis and null points (two magnetic separatrix points) is developed. The six degrees of freedom due to designated positions of the magnetic axis and null points is possible by using six poloidal field coil currents. Stable steady state tokamak plasma equilibria are calculated along with the MHD (magnetohydrodynamic) potential energy. Assuming an RF heating plasma, the plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by heating. These calculations show that, under sufficient heating, the pressure driven current in tokamak plasmas form a current hole which minimizes the MHD potential energy. (author)
Physical design of MW-class steady-state spherical tokamak, QUEST
QUEST (R=0.68 m, a=0.4 m) focuses on the steady state operation of the spherical tokamak (ST) by controlled PWI and electron Bernstain wave (EBW) current drive (CD). The QUEST project will be developed along two phases, phase I: steady state operation with plasma current, Ip=20-30 kA on open divertor configuration and phase II: steady state operation with Ip = 100 kA and β of 10% in short pulse on closed divertor configuration. Feasibility of the missions on QUEST was investigated and the suitable machine size of QUEST was decided based on the physical view of plasma parameters. Electron Bernstein wave (EBW) current drive are planned to establish the maintenance of plasma current in steady state. Mode conversion efficiency to EBW was calculated and the conversion of 95% will be expected. A new type antenna for QUEST has been fabricated to excite EBW effectively. The situation of heat and particle handling is challenging, and W and high temperature wall is adopted. The start-up scenario of plasma current was investigated based on the driven current by energetic electron and the most favorable magnetic configuration for start-up is proposed. (author)
Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime
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.
Investigation of component failure rates for pulsed versus steady state tokamak operation
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
Development of the ITER Advanced Steady State and Hybrid Scenarios
Full discharge simulations are performed to examine the plasma current rampup, flattop and rampdown phases self-consistently with the poloidal field (PF) coils and their limitations, plasma transport evolution, and heating/current drive (H/CD) sources. Steady state scenarios are found that obtain 100% non-inductive current with Ip = 7.3-10.0 MA, ΒN ∼ 2.5 for H98 = 1.6, Q's range from 3 to 6, n/nGr = 0.75-1.0, and NB, IC, EC, and LH source have been examined. The scenarios remain within CS/PF coil limits by advancing the pre-magnetization by 40 Wb. Hybrid scenarios have been identified with 35-40% non-inductive current for Ip = 12.5 MA, H98 ∼ 1.25, with q(0) reaching 1 at or after the end of rampup. The equilibrium operating space for the hybrid shows a large range of scenarios can be accommodated, and access 925-1300 s flattop burn durations.
A comparison of pulsed and steady-state tokamak reactor burn cycles. Pt. 2
Pulsed operation of a tokamak reactor imposes cost penalties due to such problems as mechanical fatigue and the need to periodically transfer large amounts of energy to various reactor components. This study focuses on lifetime limitations and capital costs of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas include: fatigue in pulsed poloidal field coils; out-of-plane bending fatigue in toroidal field coils; electric power supply costs; and noninductive current driver costs. A capital cost comparison is made for tokamak reactors operating under the four distinct operating cycles which have been proposed. Since high availability and a low cost of energy will be mandatory for a commercial fusion reactor, we can characterize improvements in physics and technology which will help achieve these goals for different burn cycles. A key conclusion is that steady-state operation is likely to result in the least expensive tokamak reactor (perhaps 20% cheaper than the best pulsed reactor), provided noninductive current drive efficiency can be increased roughly four-fold over present-day experimental results. (orig.)
Steady-state resistive toroidal-field coils for tokamak reactors
If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m2, minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems
Recent progress towards steady state tokamak operation with improved confinement in JT-60U
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)
Recent advancement in research and planning toward high beta steady state operation in KSTAR
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
Projections for a steady-state tokamak reactor based on ITER
The extensions of the physics and engineering guidelines for the ITER device needed for acceptable operating points for a steady-state tokamak power reactor are examined. Non-inductive current drive is provided in steady state by high-energy neutral-beam injection in the plasma core, lower-hybrid slow waves in the outer regions of the plasma and (30%) bootstrap current. Three different levels of extension of the ITER physics/engineering guide-lines, with differing assumptions on the possible plasma beta, elongation and aspect ratio, are considered for power-reactor applications. Plasma gain, Qp = fusion power/input power, in excess of 20 and average neutron wall fluxes from 2.3 to 3.6 MW/m2 are predicted in devices with major radii varying from 7.0 to 6.0 m and aspect ratios from 2.9 to 4. 3. Peak divertor heat fluxes range up to 12.2 MW/m2 which is somewhat higher than the current ITER design limit of 10 MW/m2 with a magnetically swept divertor. These designs were selected on the basis of improvements in physics/engineering consistent with time scales for development of future reactors. The design re-optimization on the basis of cost-of-electricity (COE) was then examined using a reactor systems model. This analysis generally verified the original estimates for the required extensions of the ITER guidelines. Cost of electricity is projected to be less than 66 mills/kWeh in all of the configurations. The smallest reactor, which has the largest neutron wall flux and mass power density, yields the lowest COE, 56 mills/kWeh. 37 refs., 8 figs., 11 tabs
Challenges to radiative divertor/mantle operations in advanced, steady-state scenarios
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
High Internal Inductance for High βN Steady-State Tokamak Operation
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.
Advanced tokamak burning plasma experiment
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)
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
Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors
Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles
Diagnostics and control for the steady state and pulsed tokamak DEMO
Orsitto, F. P.; Villari, R.; Moro, F.; Todd, T. N.; Lilley, S.; Jenkins, I.; Felton, R.; Biel, W.; Silva, A.; Scholz, M.; Rzadkiewicz, J.; Duran, I.; Tardocchi, M.; Gorini, G.; Morlock, C.; Federici, G.; Litnovsky, A.
2016-02-01
The present paper is devoted to a first assessment of the DEMO diagnostics systems and controls in the context of pulsed and steady state reactor design under study in Europe. In particular, the main arguments treated are: (i) The quantities to be measured in DEMO and the requirements for the measurements; (ii) the present capability of the diagnostic and control technology, determining the most urgent gaps, and (iii) the program and strategy of the research and development (R&D) needed to fill the gaps. Burn control, magnetohydrodynamic stability, and basic machine protection require improvements to the ITER technology, and moderated efforts in R&D can be dedicated to infrared diagnostics (reflectometry, electron cyclotron emission, polarimetry) and neutron diagnostics. Metallic Hall sensors appear to be a promising candidate for magnetic measurements in the high neutron fluence and long/steady state discharges of DEMO.
Advanced tokamak operating modes in TPX and ITER
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
Major progress on tore supra toward steady state operation of tokamaks
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)
Comparative study of pulsed and steady-state tokamak reactor burn cycles
Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles
Design constraints for rf-driven steady-state tokamak reactors
Plasma current density profiles are computed due to electron Landau damping of lower hybrid waves launched into model tokamak density and temperature profiles. The total current and current profile shape are chosen consistent with magnetohydrodynamic equilibrium for a variety of temperature and density distributions and plasma beta values. Surface current equilibria appear attractive and are accessible to waves with n/sub z/ as low as 1.2. By suitably choosing the spectrum location and width it is possible to drive the 9.8 MA current of a 7.0-m reactor with as little as 2.8% of the fusion power recirculated as rf input from the waveguides
Woolley, Robert D.
1998-01-01
A method and apparatus for the steady-state measurement of poloidal magnetic field near a tokamak plasma, where the tokamak is configured with respect to a cylindrical coordinate system having z, phi (toroidal), and r axes. The method is based on combining the two magnetic field principles of induction and torque. The apparatus includes a rotor assembly having a pair of inductive magnetic field pickup coils which are concentrically mounted, orthogonally oriented in the r and z directions, and coupled to remotely located electronics which include electronic integrators for determining magnetic field changes. The rotor assembly includes an axle oriented in the toroidal direction, with the axle mounted on pivot support brackets which in turn are mounted on a baseplate. First and second springs are located between the baseplate and the rotor assembly restricting rotation of the rotor assembly about its axle, the second spring providing a constant tensile preload in the first spring. A strain gauge is mounted on the first spring, and electronic means to continually monitor strain gauge resistance variations is provided. Electronic means for providing a known current pulse waveform to be periodically injected into each coil to create a time-varying torque on the rotor assembly in the toroidal direction causes mechanical strain variations proportional to the torque in the mounting means and springs so that strain gauge measurement of the variation provides periodic magnetic field measurements independent of the magnetic field measured by the electronic integrators.
Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations
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
Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations
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.
Update to advanced neutron source steady-state thermal-hydraulic report
This report is intended to be a supplement to ORNL/TM-12398, Steady-State Thermal-Hydraulic Design Analysis of the Advanced Neutron Source Reactor. It updates the core thermal-hydrualic design to the latest three-element configuration and also provides the most recent information on the thermal-hydraulic statistical uncertainty analysis. In addition, it includes calculations of beam tube cooling and control rod lift forces, which were not addressed in the initial report. This report describes work that is a snapshot in time as it stood at the end of the project. The three-element core calculations include a description of changes made to the overall coolant system; however, most of the analysis is focused on fuel loading thermal-hydraulic calculations. This analysis uses updated uncertainty values and indicates that a two-dimensional fuel grading in the three-element core would still be necessary to meet the desired operating and safety criteria. Analysis of cooling in the reflector tank examines various cooling options for the reflector tank components. This work investigated multiple forced convection designs as well as natural convection cooling requirements. Lift forces on the inner control rods caused by the upward coolant flow were also examined. Initial control rod designs were such that a sheared control rod would tend to lift because of flow forces. Design changes were recommended that would eliminate this issue. They included geometry changes to the inner control rod cooling channels, changes to the orificing in the central hole region, and reduction of inner control rod coolant velocity
Experimental and theoretical basis for advanced tokamaks
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
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)
ADX - Advanced Divertor and RF Tokamak Experiment
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.
Advancing the detection of steady-state visual evoked potentials in brain–computer interfaces
Abu-Alqumsan, Mohammad; Peer, Angelika
2016-06-01
Objective. Spatial filtering has proved to be a powerful pre-processing step in detection of steady-state visual evoked potentials and boosted typical detection rates both in offline analysis and online SSVEP-based brain–computer interface applications. State-of-the-art detection methods and the spatial filters used thereby share many common foundations as they all build upon the second order statistics of the acquired Electroencephalographic (EEG) data, that is, its spatial autocovariance and cross-covariance with what is assumed to be a pure SSVEP response. The present study aims at highlighting the similarities and differences between these methods. Approach. We consider the canonical correlation analysis (CCA) method as a basis for the theoretical and empirical (with real EEG data) analysis of the state-of-the-art detection methods and the spatial filters used thereby. We build upon the findings of this analysis and prior research and propose a new detection method (CVARS) that combines the power of the canonical variates and that of the autoregressive spectral analysis in estimating the signal and noise power levels. Main results. We found that the multivariate synchronization index method and the maximum contrast combination method are variations of the CCA method. All three methods were found to provide relatively unreliable detections in low signal-to-noise ratio (SNR) regimes. CVARS and the minimum energy combination methods were found to provide better estimates for different SNR levels. Significance. Our theoretical and empirical results demonstrate that the proposed CVARS method outperforms other state-of-the-art detection methods when used in an unsupervised fashion. Furthermore, when used in a supervised fashion, a linear classifier learned from a short training session is able to estimate the hidden user intention, including the idle state (when the user is not attending to any stimulus), rapidly, accurately and reliably.
MHD stability of advanced tokamak scenarios
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
Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor
The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan
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
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
Devriese, Lot A; Koch, Kevin M; Mergui-Roelvink, Marja; Matthys, Gemma M; Ma, Wen Wee; Robidoux, Andre; Stephenson, Joe J; Chu, Quincy S C; Orford, Keith W; Cartee, Leanne; Botbyl, Jeff; Arya, Nikita; Schellens, Jan H M
2014-01-01
AIM: To quantify the effect of food on the systemic exposure of lapatinib at steady state when administered 1 h before and after meals, and to observe the safety and tolerability of lapatinib under these conditions in patients with advanced solid tumours. METHODS: This was a three-treatment, randomi
Physics design of advanced steady-state tokamak reactor A-SSTR2
Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (Rp=6.2m, ap=1.5m, Ip=12MA) has a fusion power of 4GW with βN=4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)
Full text: EAST will be one of the world's first magnetic confinement devices that must address Plasma- Wall Interaction (PWI) issues facing high power steady-state operations. EAST has recently significantly augmented its RF heating capabilities up to 10 MW, including LHCD and ICRH. It has also undertaken an extensive upgrade during the recent shutdown to replace the carbon tiles on the main chamber wall and divertor surface by the Mo tiles, except those near the strike points, allowing baking up to 250 deg C, with active water cooling. The divertor titles will further be upgraded to monoblock Tungsten, as to be used in ITER, to address PWI issues for ITER and DEMO. EAST demonstrated long pulse operation over 100 s, entirely driven by LHCD during the last experimental campaign. In order to achieve this, the following major means were applied to EAST to actively control PWI interactions: 1. Active divertor pumping using an in-vessel large capacity cryopump for facilitating density control. 2. Advanced wall conditioning with Lithium (Li) evaporation and real-time, in-situ Li powder injection for controlling neutral recycling. 3. Localized divertor gas puffing for reducing peak heat fluxes near the strike points. 4. Strike point sweeping to spread the heat loads on the divertor target plates. In addition, highly radiative impurity Ar was injected into the divertor to further reduce the peak divertor heat fluxes and mitigate the in-out divertor plasma asymmetries in EAST. Despite the injection of Ar, Zeff in the core plasma was little affected, suggesting strong divertor screening. Ar seeding has also been explored in the newly achieved H-modes in EAST, significantly increasing the frequency and decreasing the amplitude of ELMs, thus reducing the particle and heat loads on the divertor target plates. These first results are very promising, and will further be investigated in EAST for high power, long pulse operations. EAST has now just started a new experimental
The methodology used to develop the LMFBR carbide/nitride fuels code, LIFE4-CN, is described in detail along with some subtleties encountered in code development. Fuel primary and steady-state thermal creep have been used as an example to illustrate the need for physical modeling and the need to recognize the importance of the materials characteristics. A self-consistent strategy for LIFE4-CN verification against irradiation data has been outlined with emphasis on the establishment of the gross uncertainty bands. These gross uncertainty bands can be used as an objective measure to gauge the overall success of the code predictions. Preliminary code predictions for sample steady-state and transient cases are given
The physics efficiency of current drive (γB ∝ ne I0 R0/PCD), including the bootstrap effect, needs to exceed certain goals in order to provide economical steady state operation compared to pulsed power plants. The goal for γB depends not only on engineering performance of the current drive system, but also on normalized beta and the effective safety factor of the achievable MHD equilibrium
This meeting has provided an appropriate forum to discuss current issues covering a wide range of technical topics related to the steady state operation issues and also to encourage forecast of the ITER performances. The technical meeting includes invited and contributed papers. The topics that have been dealt with are: 1) Superconducting devices (ITER, KSTAR, Tore-Supra, HT-7U, EAST, LHD, Wendelstein-7-X,...); 2) Long-pulse operation and advanced tokamak physics; 3) steady state fusion technologies; 4) Long pulse heating and current drive; 5) Particle control and power exhaust, and 6) ITER-related research and development issues. This document gathers the abstracts
Modelling of pulsed and steady-state DEMO scenarios
Giruzzi, G.; Artaud, J. F.; Baruzzo, M.; Bolzonella, T.; Fable, E.; Garzotti, L.; Ivanova-Stanik, I.; Kemp, R.; King, D. B.; Schneider, M.; Stankiewicz, R.; Stępniewski, W.; Vincenzi, P.; Ward, D.; Zagórski, R.
2015-07-01
Scenario modelling for the demonstration fusion reactor (DEMO) has been carried out using a variety of simulation codes. Two DEMO concepts have been analysed: a pulsed tokamak, characterized by rather conventional physics and technology assumptions (DEMO1) and a steady-state tokamak, with moderately advanced physics and technology assumptions (DEMO2). Sensitivity to impurity concentrations, radiation, and heat transport models has been investigated. For DEMO2, the impact of current driven non-inductively by neutral beams has been studied by full Monte Carlo simulations of the fast ion distribution. The results obtained are a part of a more extensive research and development (R&D) effort carried out in the EU in order to develop a viable option for a DEMO reactor, to be adopted after ITER for fusion energy research.
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
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)
Simonin, A.; Achard, Jocelyn; Achkasov, K.; Bechu, S.; Baudouin, C.; Baulaigue, O.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; de Esch, H. P. L.; Fiorucci, D.; Fubiani, G.; Furno, I.; Futtersack, R.; Garibaldi, P.; Gicquel, A.; Grand, C.; Guittienne, Ph.; Hagelaar, G.; Howling, A.; Jacquier, R.; Kirkpatrick, M. J.; Lemoine, D.; Lepetit, B.; Minea, T.; Odic, E.; Revel, A.; Soliman, B. A.; Teste, P.
2015-11-01
Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of neutral beam (NB) system for the future fusion reactor (DEMO Tokamak) has been underway between several laboratories in Europe. The specifications required to operate a NB system on DEMO are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (η > 60%), high availability and reliability. To this aim, a novel NB concept based on the photodetachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photoneutralizer where a high power photon flux (~3 MW) generated within a Fabry-Perot cavity will overlap, cross and partially photodetach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that such a photoneutralized based NB system would have the capability to provide several tens of MW of D0 per beam line with a wall-plug efficiency higher than 60%. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e. negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.
INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
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
INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
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.
Real-Time Profile Control for Advanced Tokamak Operation
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
Advanced fusion technologies developed for JT-60 superconducting tokamak
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)
The optimized li advanced tokamak scenario with high bootstrap current fraction
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
Multiple Steady States in Distillation
Bekiaris, Nikolaos
1995-01-01
We study multiple steady states in distillation. We first analyze the simplest case of ternary homogeneous azeotropic mixtures. We show that in the case of infinite reflux and an infinite number of trays (∞/∞ case) one can construct bifurcation diagrams on physical grounds with the distillate flow as the bifurcation parameter. Multiple steady states exist when the distillate flow varies non-monotonically along the continuation path of the bifurcation diagram. We derive a necessary and suffici...
Status of and prospects for advanced Tokamak regimes from multi-machine comparisons
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.)
Steady-State Process Modelling
Cameron, Ian; Gani, Rafiqul
illustrate the “equation oriented” approach as well as the “sequential modular” approach to solving complex flowsheets for steady state applications. The applications include the Williams-Otto plant, the hydrodealkylation (HDA) of toluene, conversion of ethylene to ethanol and a bio-ethanol process.......This chapter covers the basic principles of steady state modelling and simulation using a number of case studies. Two principal approaches are illustrated that develop the unit operation models from first principles as well as through application of standard flowsheet simulators. The approaches...
Owens, J. A.
1982-01-01
Options for faculty utilization in a steady state are examined, with consideration for their economy or ability to increase turnover or flexibility: early retirement, part retirement, retraining, exchange with other institutions or industry, and fixed-term appointments or lecturer positions. (MSE)
Steady-State Process Modelling
Cameron, Ian; Gani, Rafiqul
2011-01-01
illustrate the “equation oriented” approach as well as the “sequential modular” approach to solving complex flowsheets for steady state applications. The applications include the Williams-Otto plant, the hydrodealkylation (HDA) of toluene, conversion of ethylene to ethanol and a bio-ethanol process....
Advanced tokamak physics experiments on DIII-D
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.
Advanced tokamak physics experiments on DIII-D
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
Steady-state compact neutron sources with HTS magnets
Full text of publication follows. Recent advantages in the development of high temperature superconductors (HTS), and encouraging results of first tests of HTS coils on a tokamak [1], open new prospects for compact high field TF magnets for Spherical Tokamaks (STs). High β (ratio of the plasma pressure to magnetic pressure) values have been achieved in STs, which opens a path to compact Fusion devices, as the Fusion power is proportional to β2Bt4V. To make advantages of high β in compact STs, the toroidal field should be maximised, which is challenging, and all present STs operate at fields < 1 T. The favourable dependence of confinement on Bt recently found in STs [2] may allow enhanced performance in high-field STs, also encouraging increase in Bt. We investigate feasibility of HTS magnets in next-step STs and compare such designs with proposed conventional aspect ratio designs with HTS magnets (VECTOR, VULCAN etc). Main issues are: - the capital cost (will the use of HTS increase the capital cost?); - running cost (can the use of HTS reduce the running cost?); - will increase in the field in STs easy requirements on current drive?; - how much use of HTS will affect the size (e.g. the cost) of a neutron source (divertor, blanket maintenance options, shielding etc.)? Several physics aspects of a low- and medium-power steady-state neutron source will be discussed. These include fast particle and alpha particle losses, effect of increase in Bt on micro-stability etc. The demonstration of reliable steady state operations in a compact ST even at the level of a few MW Fusion output (which easy application of HTS) as a first step will significantly advance not only the mainstream Fusion for Energy research, but also the commercial exploitation of Fusion Power. [1] M Gryaznevich et al., 'Progress in applications of High Temperature Superconductor in Tokamak Magnets', Fusion Engineering and Design, accepted for publication, (2013). [2] M. Valovic et al, Nucl
Steady-state eternal inflation
Since the advent of inflation, several theorems have been proven suggesting that although inflation can (and generically does) continue eternally into the future, it cannot be extended eternally into the past to create a 'steady-state' model with no initial time. Here we provide a construction that circumvents these theorems and allows a self-consistent, geodesically complete, and physically sensible steady-state eternally inflating universe, based on the flat slicing of de Sitter space. This construction could be used as the background spacetime for creation events that form big-bang-like regions, and hence could form the basis for a cosmology that is compatible with observations and yet which avoids an initial singularity or beginning of time
Steady state neutral beam injector
Learning from operational reliability of neutral beam injectors in particular and various heating schemes including RF in general on TFTR, JET, JT-60, it has become clear that neutral beam injectors may find a greater role assigned to them for maintaining the plasma in steady state devices under construction. Many technological solutions, integrated in the present day generation of injectors have given rise to capability of producing multimegawatt power at many tens of kV. They have already operated for integrated time >105 S without deterioration in the performance. However, a new generation of injectors for steady state devices have to address to some basic issues. They stem from material erosion under particle bombardment, heat transfer > 10 MW/m2, frequent regeneration of cryopanels, inertial power supplies, data acquisition and control of large volume of data. Some of these engineering issues have been addressed to in the proposed neutral beam injector for SST-1 at our institute; the remaining shall have to wait for the inputs of the database generated from the actual experience with steady state injectors. (author)
Tokamak advanced pump limiter experiments and analysis
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
Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER
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
Control of the Resistive Wall Mode in Advanced Tokamak Plasmas on DIII-D
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
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
Transient heat transport studies in JET conventional and advanced tokamak plasmas
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)
Summary discussion: An integrated advanced tokamak reactor
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
Inconsistencies in steady state thermodynamics
Dickman, Ronald; Motai, Ricardo
2014-03-01
We address the issue of extending thermodynamics to nonequilibrium steady states. Using driven stochastic lattice gases, we ask whether consistent definitions of an effective chemical potential μ, and an effective temperature Te, are possible. These quantities are determined via zero-flux conditions of particles and energy between the driven system and a reservoir. For the models considered here, the fluxes are given in terms of certain stationary average densities, eliminating the need to perturb the system by actually exchanging particles; μ and Te are thereby obtained via open-circuit measurements, using a virtual reservoir. In the lattice gas with nearest-neighbor exclusion, temperature is not relevant, and we find that the effective chemical potential, a function of density and drive strength, satisfies the zeroth law, and correctly predicts the densities of coexisting systems. In the Katz-Lebowitz-Spohn driven lattice gas, both μ and Te need to be defined. We show analytically that the zeroth law is violated, and determine the size of the violations numerically. Our results highlight a fundamental inconsistency in the extension of thermodynamics to nonequilibrium steady states. Research supported by CNPq, Brazil.
Collective effects and self-consistent energetic particle dynamics in advanced tokamaks
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)
Accelerator based steady state neutron source
Using high current, cw linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the accelerator based neutron research facility (ABNR) would initially achieve the 1016 n/cm2s thermal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of $300-450 M is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source is most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc., with the development of a multibeam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs
An accelerator based steady state neutron source
Using high current, cw linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the accelerator based neutron research facility (ABNR) would initially achieve the 1016 n/cm2 s themal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of Dollar 300-450 is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source in most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc. With the development of a multibeam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs. (orig.)
The technology and science of steady-state operation in magnetically confined plasmas
The steady-state operation of magnetically confined fusion plasmas is considered as one of the 'grand challenges' of future decades, if not the ultimate goal of the research and development activities towards a new source of energy. Reaching such a goal requires the high-level integration of both science and technology aspects of magnetic fusion into self-consistent plasma regimes in fusion-grade devices. On the physics side, the first constraint addresses the magnetic confinement itself which must be made persistent. This means to either rely on intrinsically steady-state configurations, like the stellarator one, or turn the inductively driven tokamak configuration into a fully non-inductive one, through a mix of additional current sources. The low efficiency of the external current drive methods and the necessity to minimize the re-circulating power claim for a current mix strongly weighted by the internal 'pressure driven' bootstrap current, itself strongly sensitive to the heat and particle transport properties of the plasma. A virtuous circle may form as the heat and particle transport properties are themselves sensitive to the current profile conditions. Note that several other factors, e.g. plasma rotation profile, magneto-hydro-dynamics activity, also influence the equilibrium state. In the present tokamak devices, several examples of such 'advanced tokamak' physics research demonstrate the feasibility of steady-state regimes, though with a number of open questions still under investigation. The modelling activity also progresses quite fast in this domain and supports understanding and extrapolation. This high level of physics sophistication of the plasma scenario however needs to be combined with steady-state technological constraints. The technology constraints for steady-state operation are basically twofold: the specific technologies required to reach the steady-state plasma conditions and the generic technologies linked to the long pulse operation of a
Development of burning plasma and advanced scenarios in the DIII-D tokamak
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
Progress towards high-performance, steady-state spherical torus.
Lee, S.G (Korea Basic Science Institute, Taejon, Republic of Korea); Kugel, W. (Princeton University, NJ); Efthimion, P. C. (Princeton University, NJ); Kissick, M. W. (University of Wisconsin, WI); Bourdelle, C. (CEA Cadarache, France); Kim, J.H (Korea Advanced Institute of Science and Technology, Taejon, Republic of Korea); Gray, T. (Princeton University, NJ); Garstka, G. D. (University of Wisconsin, WI); Fonck, R. J. (University of Wisconsin, WI); Doerner, R. (University of California, San Diego, CA); Diem, S.J. (University of Wisconsin, WI); Pacella, D. (ENEA, Frascati, Italy); Nishino, N. (Hiroshima University, Hiroshima, Japan); Ferron, J. R. (General Atomics, San Diego, CA); Skinner, C. H. (Princeton University, NJ); Stutman, D. (Johns Hopkins University, Baltimore, MD); Soukhanovskii, V. (Princeton University, NJ); Choe, W. (Korea Advanced Institute of Science and Technology, Taejon, Republic of Korea); Chrzanowski, J. (Princeton University, NJ); Mau, T.K. (University of California, San Diego, CA); Bell, Michael G. (Princeton University, NJ); Raman, R. (University of Washington, Seattle, WA); Peng, Y-K. M. (Oak Ridge National Laboratory, Oak Ridge, TN); Ono, M. (Princeton University, NJ); Park, W. (Princeton University, NJ); Hoffman, D. (Princeton University, NJ); Maqueda, R. (Los Alamos National Laboratory, Los Alamos, NM); Kaye, S. M. (Princeton University, NJ); Kaita, R. (Princeton University, NJ); Jarboe, T.R. (University of Washington, Seattle, WA); Hill, K.W. (Princeton University, NJ); Heidbrink, W. (University of California, Irvine, CA); Spaleta, J. (Princeton University, NJ); Sontag, A.C (University of Wisconsin, WI); Seraydarian, R. (University of California, San Diego, CA); Schooff, R.J. (University of Wisconsin, WI); Sabbagh, S.A. (Columbia University, New York, NY); Menard, J. (Princeton University, NJ); Mazzucato, E. (Princeton University, NJ); Lee, K. (University of California, Davis, CA); LeBlanc, B. (Princeton University, NJ); Probert, P. H. (University of Wisconsin, WI); Blanchard, W. (Princeton University, NJ); Wampler, William R.; Swain, D. W. (Oak Ridge National Laboratory, Oak Ridge, TN); Ryan, P.M. (Oak Ridge National Laboratory, Oak Ridge, TN); Rosenberg, A. (Princeton University, NJ); Ramakrishnan, S. (Princeton University, NJ); Phillips, C.K. (Princeton University, NJ); Park, H.K. (Princeton University, NJ); Roquemore, A. L. (Princeton University, NJ); Paoletti, F. (Columbia University, New York, NY); Medley, S. S. (Princeton University, NJ); Fredrickson, E. D. (Princeton University, NJ); Kessel, C. E. (Princeton University, NJ); Stevenson, T. (Princeton University, NJ); Darrow, D. S. (Princeton University, NJ); Majeski, R. (Princeton University, NJ); Bitter, M. (Princeton University, NJ); Neumeyer, C. (Princeton University, NJ); Nelson, B.A. (University of Washington, Seattle, WA); Paul, S. F. (Princeton University, NJ); Manickam, J. (Princeton University, NJ); Ostrander, C. N. (University of Wisconsin, WI); Mueller, D. (Princeton University, NJ); Lewicki, B.T (University of Wisconsin, WI); Luckhardt, S. (University of California, San Diego, CA); Johnson, D.W. (Princeton University, NJ); Grisham, L.R. (Princeton University, NJ); Kubota, Shigeru (University of California, Los Angeles, CA); Gates, D.A. (Princeton University, NJ); Bush, C. (Oak Ridge National Laboratory, Oak Ridge, TN); Synakowski, E.J. (Princeton University, NJ); Schaffer, M. (General Atomics, San Diego, CA); Boedo, J. (University of California, San Diego, CA); Maingi, R. (Oak Ridge National Laboratory, Oak Ridge, TN); Redi, M. (Princeton University, NJ); Pinsker, R. (General Atomics, San Diego, CA); Bigelow, T. (Oak Ridge National Laboratory, Oak Ridge, TN); Bell, R. E. (Princeton University, NJ)
2004-06-01
Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta ({beta}), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values {beta}{sub T} of up to 35% with a near unity central {beta}{sub T} have been obtained. NSTX will be exploring advanced regimes where {beta}{sub T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction ({approx}60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX
Progress Towards High Performance, Steady-state Spherical Torus
M. Ono; M.G. Bell; R.E. Bell; T. Bigelow; M. Bitter; W. Blanchard; J. Boedo; C. Bourdelle; C. Bush; W. Choe; J. Chrzanowski; D.S. Darrow; S.J. Diem; R. Doerner; P.C. Efthimion; J.R. Ferron; R.J. Fonck; E.D. Fredrickson; G.D. Garstka; D.A. Gates; T. Gray; L.R. Grisham; W. Heidbrink; K.W. Hill; D. Hoffman; T.R. Jarboe; D.W. Johnson; R. Kaita; S.M. Kaye; C. Kessel; J.H. Kim; M.W. Kissick; S. Kubota; H.W. Kugel; B.P. LeBlanc; K. Lee; S.G. Lee; B.T. Lewicki; S. Luckhardt; R. Maingi; R. Majeski; J. Manickam; R. Maqueda; T.K. Mau; E. Mazzucato; S.S. Medley; J. Menard; D. Mueller; B.A. Nelson; C. Neumeyer; N. Nishino; C.N. Ostrander; D. Pacella; F. Paoletti; H.K. Park; W. Park; S.F. Paul; Y.-K. M. Peng; C.K. Phillips; R. Pinsker; P.H. Probert; S. Ramakrishnan; R. Raman; M. Redi; A.L. Roquemore; A. Rosenberg; P.M. Ryan; S.A. Sabbagh; M. Schaffer; R.J. Schooff; R. Seraydarian; C.H. Skinner; A.C. Sontag; V. Soukhanovskii; J. Spaleta; T. Stevenson; D. Stutman; D.W. Swain; E. Synakowski; Y. Takase; X. Tang; G. Taylor; J. Timberlake; K.L. Tritz; E.A. Unterberg; A. Von Halle; J. Wilgen; M. Williams; J.R. Wilson; X. Xu; S.J. Zweben; R. Akers; R.E. Barry; P. Beiersdorfer; J.M. Bialek; B. Blagojevic; P.T. Bonoli; M.D. Carter; W. Davis; B. Deng; L. Dudek; J. Egedal; R. Ellis; M. Finkenthal; J. Foley; E. Fredd; A. Glasser; T. Gibney; M. Gilmore; R.J. Goldston; R.E. Hatcher; R.J. Hawryluk; W. Houlberg; R. Harvey; S.C. Jardin; J.C. Hosea; H. Ji; M. Kalish; J. Lowrance; L.L. Lao; F.M. Levinton; N.C. Luhmann; R. Marsala; D. Mastravito; M.M. Menon; O. Mitarai; M. Nagata; G. Oliaro; R. Parsells; T. Peebles; B. Peneflor; D. Piglowski; G.D. Porter; A.K. Ram; M. Rensink; G. Rewoldt; P. Roney; K. Shaing; S. Shiraiwa; P. Sichta; D. Stotler; B.C. Stratton; R. Vero; W.R. Wampler; G.A. Wurden
2003-10-02
Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been
Progress towards high-performance, steady-state spherical torus
Ono, M.; Bell, M. G.; Bell, R. E.; Bigelow, T.; Bitter, M.; Blanchard, W.; Boedo, J.; Bourdelle, C.; Bush, C.; Choe, W.; Chrzanowski, J.; Darrow, D. S.; Diem, S. J.; Doerner, R.; Efthimion, P. C.; Ferron, J. R.; Fonck, R. J.; Fredrickson, E. D.; Garstka, G. D.; Gates, D A; Gray, T.; Grisham, L. R.; Heidbrink, W.; Hill, K. W.; Hoffman, D.; Jarboe, T. R.; Johnson, D. W.; Kaita, R.; Kaye, S. M.; Kessel, C.; Kim, J. H.; Kissick, M. W.; Kubota, S.; Kugel, H. W.; LeBlanc, B. P.; Lee, K.; Lee, S. G.; Lewicki, B. T.; Luckhardt, S.; Maingi, R.; Majeski, R.; Manickam, J.; Maqueda, R.; Mau, T. K.; Mazzucato, E.; Medley, S. S.; Menard, J.; Mueller, D.; Nelson, B. A.; Neumeyer, C.; Nishino, N.; Ostrander, C. N.; Pacella, D.; Paoletti, F.; Park, H. K.; Park, W.; Paul, S. F.; Peng, Y-K M.; Phillips, C. K.; Pinsker, R.; Probert, P. H.; Ramakrishnan, S.; Raman, R.; Redi, M.; Roquemore, A. L.; Rosenberg, A.; Ryan, P. M.; Sabbagh, S. A.; Schaffer, M.; Schooff, R. J.; Seraydarian, R.; Skinner, C. H.; Sontag, A. C.; Soukhanovskii, V.; Spaleta, J.; Stevenson, T.; Stutman, D.; Swain, D. W.; Synakowski, E.; Takase, Y.; Tang, X.; Taylor, G.; Timberlake, J.; Tritz, K. L.; Unterberg, E. A.; Halle, A. Von.; Wilgen, J.; Williams, M.; Wilson, J. R.; Xu, X.; Zweben, S. J.; Akers, R.; Barry, R. E.; Beiersdorfer, P.; Bialek, J. M.; Blagojevic, B.; Bonoli, P. T.; Carter, M. D.; Davis, W.; Deng, B.; Dudek, L.; Egedal, J.; Ellis, R.; Finkenthal, M.; Foley, J.; Fredd, E.; Glasser, A.; Gibney, T.; Gilmore, M.; Goldston, R. J.; Hatcher, R. E.; Hawryluk, R. J.; Houlberg, W.; Harvey, R.; Jardin, S. C.; Hosea, J. C.; Ji, H.; Kalish, M.; Lowrance, J.; Lao, L. L.; Levinton, F. M.; Luhmann, N. C.; Marsala, R.; Mastravito, D.; Menon, M. M.; Mitarai, O.; Nagata, M.; Oliaro, G.; Parsells, R.; Peebles, T.; Peneflor, B.; Piglowski, D.; Porter, G. D.; Ram, A. K.; Rensink, M.; Rewoldt, G.; Robinson, J.; Roney, P.; Shaing, K.; Shiraiwa, S.; Sichta, P.; Stotler, D.; Stratton, B. C.; Vero, R.; Wampler, W. R.; Wurden, G. A.
2003-12-01
Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (β), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values β_{T} of up to 35% with a near unity central β_{T} have been obtained. NSTX will be exploring advanced regimes where β_{T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~ 60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to
Progress Towards High-Performance, Steady-State Spherical Torus
Lawrence Livermore National Laboratory
2004-01-04
Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta ({beta}), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values {beta}{sub T} of up to 35% with a near unity central {beta}{sub T} have been obtained. NSTX will be exploring advanced regimes where {beta}{sub T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction ({approx}60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fastwave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX
Progress Towards High Performance, Steady-state Spherical Torus
Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction (∼60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted
Chemical reaction systems with toric steady states
Millan, Mercedes Perez; Shiu, Anne; Conradi, Carsten
2011-01-01
Mass-action chemical reaction systems are frequently used in Computational Biology. The corresponding polynomial dynamical systems are often large, consisting of tens or even hundreds of ordinary differential equations, and poorly parameterized (due to noisy measurement data and a small number of data points and repetitions). Therefore, it is often difficult to establish the existence of (positive) steady states or to determine whether more complicated phenomena such as multistationarity exist. If, however, the steady state ideal of the system is a binomial ideal, then we show that these questions can be answered easily. The focus of this work is on systems with this property, and we say that such systems have toric steady states. Our main result gives sufficient conditions for a chemical reaction system to have toric steady states. Furthermore, we analyze the capacity of such a system to exhibit positive steady states and multistationarity. Examples of systems with toric steady states include weakly-reversib...
Iler, H. Darrell; Brown, Amber; Landis, Amanda; Schimke, Greg; Peters, George
2014-01-01
A numerical analysis of the free radical addition polymerization system is described that provides those teaching polymer, physical, or advanced organic chemistry courses the opportunity to introduce students to numerical methods in the context of a simple but mathematically stiff chemical kinetic system. Numerical analysis can lead students to an…
Advanced tokamak concepts and reactor designs
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
As highly promising coolant for new generation nuclear reactors, liquid Lead-Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations - ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI - contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal-hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.
High performance advanced tokamak regimes in DIII-D for next-step experiments
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
The aim of this thesis is to give a global scope of the problem of energy transport within a thermonuclear plasma in the context of its power balance and the implications when modelling ITER operating scenarios. This is made in two phases. First, by furnishing new elements to the existing models of heat and synchrotron radiation transport in a thermonuclear plasma. Second, by applying the improved models to plasma engineering studies of ITER operating scenarios. The scenarios modelled are the steady state operating point and the transient that appears to have the biggest technological implications: the fast burn termination. The conduction-convection losses are modelled through the energy confinement time. This parameter is empirically obtained from the existing experimental data, since the underlying mechanisms are not well understood. In chapter 2 an expression for the energy confinement time is semi-analytically deduced from the Rebut-Lallia-Watkins local transport model. The current estimates of the synchrotron radiation losses are made with expressions of the dimensionless transparency factor deduced from a 0-dimensional cylindrical model proposed by Trubnikov in 1979. In chapter 3 realistic hypothesis for the cases of cylindrical and toroidal geometry are included in the model to deduce compact explicit expressions for the fast numerical computation of the synchrotron radiation losses. Numerical applications are provided for the cylindrical case. The results are checked against the existing models. In chapter 4, the nominal operating point of ITER and its thermal stability is studied by means of a 0-dimensional burn model of the thermonuclear plasma in ignition. This model is deduced by the elements furnished by the plasma particle and power balance. Possible heat overloading on the plasma facing components may provoke severe structural damage, implying potential safety problems related to tritium inventory and metal activation. In chapter 5, the assessment
Villar Colome, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Universitat Polytechnica de Catalunya (Spain)
1997-12-01
The aim of this thesis is to give a global scope of the problem of energy transport within a thermonuclear plasma in the context of its power balance and the implications when modelling ITER operating scenarios. This is made in two phases. First, by furnishing new elements to the existing models of heat and synchrotron radiation transport in a thermonuclear plasma. Second, by applying the improved models to plasma engineering studies of ITER operating scenarios. The scenarios modelled are the steady state operating point and the transient that appears to have the biggest technological implications: the fast burn termination. The conduction-convection losses are modelled through the energy confinement time. This parameter is empirically obtained from the existing experimental data, since the underlying mechanisms are not well understood. In chapter 2 an expression for the energy confinement time is semi-analytically deduced from the Rebut-Lallia-Watkins local transport model. The current estimates of the synchrotron radiation losses are made with expressions of the dimensionless transparency factor deduced from a 0-dimensional cylindrical model proposed by Trubnikov in 1979. In chapter 3 realistic hypothesis for the cases of cylindrical and toroidal geometry are included in the model to deduce compact explicit expressions for the fast numerical computation of the synchrotron radiation losses. Numerical applications are provided for the cylindrical case. The results are checked against the existing models. In chapter 4, the nominal operating point of ITER and its thermal stability is studied by means of a 0-dimensional burn model of the thermonuclear plasma in ignition. This model is deduced by the elements furnished by the plasma particle and power balance. Possible heat overloading on the plasma facing components may provoke severe structural damage, implying potential safety problems related to tritium inventory and metal activation. In chapter 5, the assessment
Study of an advanced D-T tokamak fusion reactor with compact fusion advanced rankine (CFAR) cycle
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
Stability of infernal and ballooning modes in advanced tokamak scenarios
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
DIII-D Advanced Tokamak Research Overview
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
Lithium-cooled blankets for advanced tokamaks
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
Steady State Vapor Bubble in Pool Boiling
Zou, An; Chanana, Ashish; Agrawal, Amit; Wayner, Peter C.; Maroo, Shalabh C.
2016-02-01
Boiling, a dynamic and multiscale process, has been studied for several decades; however, a comprehensive understanding of the process is still lacking. The bubble ebullition cycle, which occurs over millisecond time-span, makes it extremely challenging to study near-surface interfacial characteristics of a single bubble. Here, we create a steady-state vapor bubble that can remain stable for hours in a pool of sub-cooled water using a femtosecond laser source. The stability of the bubble allows us to measure the contact-angle and perform in-situ imaging of the contact-line region and the microlayer, on hydrophilic and hydrophobic surfaces and in both degassed and regular (with dissolved air) water. The early growth stage of vapor bubble in degassed water shows a completely wetted bubble base with the microlayer, and the bubble does not depart from the surface due to reduced liquid pressure in the microlayer. Using experimental data and numerical simulations, we obtain permissible range of maximum heat transfer coefficient possible in nucleate boiling and the width of the evaporating layer in the contact-line region. This technique of creating and measuring fundamental characteristics of a stable vapor bubble will facilitate rational design of nanostructures for boiling enhancement and advance thermal management in electronics.
Steady State Vapor Bubble in Pool Boiling.
Zou, An; Chanana, Ashish; Agrawal, Amit; Wayner, Peter C; Maroo, Shalabh C
2016-01-01
Boiling, a dynamic and multiscale process, has been studied for several decades; however, a comprehensive understanding of the process is still lacking. The bubble ebullition cycle, which occurs over millisecond time-span, makes it extremely challenging to study near-surface interfacial characteristics of a single bubble. Here, we create a steady-state vapor bubble that can remain stable for hours in a pool of sub-cooled water using a femtosecond laser source. The stability of the bubble allows us to measure the contact-angle and perform in-situ imaging of the contact-line region and the microlayer, on hydrophilic and hydrophobic surfaces and in both degassed and regular (with dissolved air) water. The early growth stage of vapor bubble in degassed water shows a completely wetted bubble base with the microlayer, and the bubble does not depart from the surface due to reduced liquid pressure in the microlayer. Using experimental data and numerical simulations, we obtain permissible range of maximum heat transfer coefficient possible in nucleate boiling and the width of the evaporating layer in the contact-line region. This technique of creating and measuring fundamental characteristics of a stable vapor bubble will facilitate rational design of nanostructures for boiling enhancement and advance thermal management in electronics. PMID:26837464
An Advanced Tokamak Fusion Nuclear Science Facility (FNSF-AT)
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.
ADX: a high field, high power density, advanced divertor and RF tokamak
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
A concept for next step advanced tokamak fusion device
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
Optimizing the beta limit in DIII-D advanced tokamak discharges
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)
Characterisation, modelling and control of advanced scenarios in the european tokamak jet
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)
Development of burning plasma and advanced scenarios in the DIII-D tokamak
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)
Energy Transport in the Steady State Plasma Sustained by DC Helicity Current Drive
K. Itoh; Itoh, S.-I.; Fukuyama, A.
1992-01-01
Steady state operation of tokamaks which is sustained by the DC helicity current drive near edge is studied. The necessary value of the current diffusivity is obtained. Relation between the current diffusivity and the thermal diffusivity, which are governed by the microscopic turbulence, indicates that this requires too large thermal transport for the parameters in present day experiments.
The SST-1 is a superconducting tokamak, which is in the commissioning phase and will soon be ready for plasma operation. The superconducting magnet system of SST-1 comprises toroidal field (TF) and poloidal field (PF) coils. The 16 TF coils are nosed and clamped towards the in-board side and are supported toroidally with the inter-coil structure at the out-board side, forming a rigid body system. The 9 PF coils are clamped on the TF coils structure. The integrated system of TF coils and PF coils forms the cold mass of 50 ton weight. This cold mass system (CMS) is freely supported on the rigid support ring at 16 locations and the support ring in turn is supported on the 8 columns of the machine support structure. This CMS is accommodated inside the high vacuum chamber (cryostat). During the operation this cold mass attains a cryogenic temperature of 4.2 K in the hostile environment of high vacuum. During the cool down, the thermal excursion of cold mass and its supporting structure generates severe frictional forces at the sliding surfaces of the support. There is a design requirement of introducing a thin layer of solid lubricant film of molybdenum disulfide having coefficient of friction 0.05 between the sliding surfaces to control the stress contribution due to the friction. To ascertain the compatibility of molybdenum disulphide (MoS2) as a solid lubricant in a high vacuum and very low temperature environment, we have carried out qualification tests on various samples and measured the coefficient of friction in both room temperature conditions and at high vacuum and after thermal shocking at a temperature of 4.2 K in a high vacuum environment to simulate the actual working condition. After successful qualification tests and process establishment, the actual components are fabricated and integrated in the cold mass support structure assembly. The design requirement and qualification tests performed at 4.2 K and room temperature as well as details about the
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.
A steady state theory for processive cellulases
Cruys-Bagger, Nicolaj; Olsen, Jens Elmerdahl; Præstgaard, Eigil;
2013-01-01
remains to be fully developed. In this paper, we suggest a deterministic kinetic model that relies on a processive set of enzyme reactions and a quasi steady-state assumption. It is shown that this approach is practicable in the sense that it leads to mathematically simple expressions for the steady-state...... rate, and only requires data from standard assay techniques as experimental input. Specifically, it is shown that the processive reaction rate at steady state may be expressed by a hyperbolic function related to the conventional Michaelis–Menten equation. The main difference is a ‘kinetic processivity....... This has significant kinetic implications, for example the maximal specific rate (Vmax/E0) for processive cellulases is much lower than the catalytic rate constant (kcat). We discuss how relationships based on this theory may be used in both comparative and mechanistic analyses of cellulases....
Multiple steady state phenomenon in martensitic transformation
无
2001-01-01
Based on the basic facts that the martensitic transformation is a physical phenomenon which occurs in non-equilibrium conditions and there exists the feedback mechanism in the martensitic transformation, the dynamical processes of the isothermal and athermal martensitic transformations were analyzed by using nonlinear theory and a bifurcation theory model was established. It is shown that a multiple steady state phenomenon can take place as austenite is cooled, and the transitions of the steady state temperature between the branches of stable steady states can be considered the transformation from austenite to martensite. This model can estimate the starting temperature of the martensitic transformation and explain some experimental features of the martensitic transformation such as the effects of cooling rate, fluctuation and austenitic grain size on the martensitic transformation.
Physics studies for steady state operation coordinated by the ITPA
Full text of publication follows. The International Tokamak Physics Activity (ITPA) aims at cooperation on an international level in development of the physics basis for burning tokamak plasmas, supporting the preparation of ITER operation, and tokamak research worldwide. The topical group on 'Integrated Operation Scenarios' coordinates research in the following 4 areas: First, IEA collaboration experiments, coordinated by the ITPA. These experiments, performed as joint experiments in several different machines, mainly concern the validation of ITER operation scenarios, including the hybrid and steady state scenarios for ITER. Specific access conditions of these two scenarios are studied together with operation close to ITER conditions. For the heating systems, specific experiments are coordinated to study the coupling of ICRH and LHCD. Secondly, modelling and benchmarking of heating systems (actuators). Benchmarking of the actuators available for heating and current drive is an important area of international collaboration. They have been performed and completed in recent years for LHCD, for ICRH and for NBCD. In particular, LHCD at high plasma density have been studied and compared to experimental data. Thirdly, the coordinated modelling of ITER scenarios. Simulations for hybrid and steady state scenarios have a particular focus on comparing code to code results (benchmarking). For hybrid scenarios the current rise phase and the current profile evolution toward q(0)=1 were modelled with various integrated modelling codes. ITER H-mode scenarios at low plasma density have been modelled showing that the burn can be sustained for > 1000 s, suitable for neutron fluence studies in ITER. The effectiveness of the ITER day-1 heating systems for obtaining steady state scenarios as well as potential heating and current drive upgrades for ITER have been evaluated. Fourth, real time control requirements. Control of burning plasma remains a focus of research
JET steady state ITB operation with active control of the pressure profile
Stationary operations have been achieved at JET in ITBs scenarios, with the discharge time limited only by plant constraints. Full current drive was obtained, all over the high performance phase, with the current density profile frozen by using Lower Hybrid current drive. For the first time a feed-back control on the total pressure and on the electron temperature profile was implemented by using respectively the Neutral Beams and the Ion Cyclotron waves. Although impurity accumulation could be a problem in steady state ITBs, these experiments bring some elements to answer to it. Tokamak operation in enhanced confinement regimes, characterized by edge and/or Internal Transport Barriers (respectively known as H-mode and ITB), is attractive as it represents an important step towards the approach of ignition conditions. Moreover, the necessity of steady state operation in a Tokamak reactor, has led to the concept of the Advanced Tokamak, in which the current density profile is no longer tied to the plasma conductivity and is non inductively driven. Since the bootstrap current is a consequence of the pressure gradient, one of the primary goal of the Advanced Tokamak studies is to maximize the bootstrap fraction, with a proper alignment, both in H mode and in ITB regimes. However, for several reasons, it is difficult to envisage an operational situation in which the bootstrap fraction is close to 100%: for instance, there are few chances of pressure or/and current profile control to optimize the MHD stability. So far, various experiments have been performed with improved confinement regimes lasting up to tens of the confinement time and up to some current relaxation times. In some experiments a large non inductive plasma current (< 75%) was obtained with about 50% from bootstrap and 25% from Neutral Beam Injection (NBI); however, no full current drive operation was achieved and, moreover, with the available heating systems, no active feedback control of the current
LIDAR Thomson scattering for advanced tokamaks. Final report
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
Steady-state spheromak reactor studies
After summarizing the essential elements of a gun-sustained spheromak, the potential for a steady-state is explored by means of a comprehensive physics/engineering/costing model. A range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported
Thermodynamics of Stability of Nonequilibrium Steady States.
Rastogi, R. P.; Shabd, Ram
1983-01-01
Presented is a concise and critical account of developments in nonequilibrium thermodynamics. The criterion for stability of nonequilibrium steady states is critically examined for consecutive and monomolecular triangular reactions, autocatalytic reactions, auto-inhibited reactions, and the Lotka-Volterra model. (JN)
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)
Steady State Analysis of Towed Marine Cables
WANG Fei; HUANG Guo-liang; DENG De-heng
2008-01-01
Efficient numerical schemes were presented for the steady state solutions of towed marine cables. For most of towed systems, the steady state problem can be resolved into two-point boundary-value problem, or initial value problem in some special cases where the initial values are available directly. A new technique was proposed and attempted to solve the two-point boundary-value problem rather than the conventional shooting method due to its algorithm complexity and low efficiency. First, the boundary conditions are transformed into a set of nonlinear governing equations about the initial values, then bisection method is employed to solve these nonlinear equations with the aid of 4th order Runge-Kutta method. In common sense, non-uniform (sheared) current is assumed, which varies in magnitude and direction with depth. The schemes are validated through the DE Zoysa's example, then several numerical examples are also presented to illustrate the numerical schemes.
Development of steady state magnetic sensor
Hara, Shigemitsu; Nakayama, Takahide [Hitachi Ltd., Tokyo (Japan); Nagashima, Akira; Kasai, Satoshi
1998-12-01
A prototype of new mechanical sensor based on the steady state electromagnetic force (J x B force) measurement has been developed and tested. The mechanical force sensor is a new type of the magnetic sensor which is available for frequencies smaller than 0.1 Hz. The prototype of the mechanical sensor has been examined, and the following results were obtained; (1) A signal was proportional to simulated force in the load cell tests. (2) A signal drift concerning the temperature was reproducible over the range of the ITER environment. (3) A signal was proportional to the magnetic field in the steady state magnetic field measurement tests. (4) A load cell linearity error did not increase significantly after irradiation of 7.2 x 10{sup 6} Gy. These results indicate that the mechanical sensor will provide the practical feasibility in the long time magnetic field measurement. (author)
Halo current diagnostic system of experimental advanced superconducting tokamak.
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
Design and construction of the KSTAR tokamak
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)
Variational methods in steady state diffusion problems
Classical variational techniques are used to obtain accurate solutions to the multigroup multiregion one dimensional steady state neutron diffusion equation. Analytic solutions are constructed for benchmark verification. Functionals with cubic trial functions and conservational lagrangian constraints are exhibited and compared with nonconservational functionals with respect to neutron balance and to relative flux and current at interfaces. Excellent agreement of the conservational functionals using cubic trial functions is obtained in comparison with analytic solutions
On Typicality in Nonequilibrium Steady States
Evans, Denis J.; Williams, Stephen R.; Searles, Debra J.; Rondoni, Lamberto
2016-08-01
From the statistical mechanical viewpoint, relaxation of macroscopic systems and response theory rest on a notion of typicality, according to which the behavior of single macroscopic objects is given by appropriate ensembles: ensemble averages of observable quantities represent the measurements performed on single objects, because " almost all" objects share the same fate. In the case of non-dissipative dynamics and relaxation toward equilibrium states, " almost all" is referred to invariant probability distributions that are absolutely continuous with respect to the Lebesgue measure. In other words, the collection of initial micro-states (single systems) that do not follow the ensemble is supposed to constitute a set of vanishing, phase space volume. This approach is problematic in the case of dissipative dynamics and relaxation to nonequilibrium steady states, because the relevant invariant distributions attribute probability 1 to sets of zero volume, while evolution commonly begins in equilibrium states, i.e., in sets of full phase space volume. We consider the relaxation of classical, thermostatted particle systems to nonequilibrium steady states. We show that the dynamical condition known as Ω T-mixing is necessary and sufficient for relaxation of ensemble averages to steady state values. Moreover, we find that the condition known as weak T-mixing applied to smooth observables is sufficient for ensemble relaxation to be independent of the initial ensemble. Lastly, we show that weak T-mixing provides a notion of typicality for dissipative dynamics that is based on the (non-invariant) Lebesgue measure, and that we call physical ergodicity.
Neutral beam injection (NBI) system is a workhorse to heat magnetically confined tokamak fusion plasma. The heart of any NBI system is an ion extractor system. Steady State Superconducting Tokamak-1 (SST-1) needs 0.5 MW of hydrogen beam power at 30 kV to raise the plasma ion temperature to ∼1 keV and 1.7 MW of hydrogen beam power at 55 kV for future upgradation. To meet this requirement, an ion extractor system consisting of three actively cooled grids has been designed, fabricated, and its performance test has been done at MARION test stand, IPP, Julich, Germany. During long pulse (14 s) operation, hydrogen ion beam of energy 31 MJ has been extracted at 41 kV. In this paper, we have presented detailed analysis of calorimetric data of actively cooled extractor grids and showed that by monitoring outlet water temperature, grid material temperature can be monitored for safe steady state operation of a NBI system. Steady state operation of NBI is the present day interest of fusion research. In the present experimental case, performance test analysis indicates that the actively cooled grids attain steady state heat removal condition and the grid material temperature rise is ∼18 deg. C and saturates after 10 s of beam pulse.
Long Pulse Operation on Tore-Supra: Towards Steady State
The experimental programme of Tore Supra is devoted to the study of technology and physics issues associated to long-duration high performance discharges. This new domain of operation requires simultaneously and in steady state: heat removal capability, particle exhaust, fully non-inductive current drive, advanced technology integration and real time plasma control. The long discharge allows for addressing new time scale physic such as the wall particle retention and erosion. Moreover, the physics of fully non-inductive discharges is full of novelty, namely: the MHD stability, the slow spontaneous oscillation of the central electron temperature or the outstanding inward particle pinch
On the optimization of a steady-state bootstrap-reactor
A commercial fusion tokamak-reactor may be economically acceptable only for low recirculating power fraction r0 ≡ PCD/Pα BS≡IBS/I > 0.9 to sustain the steady-state operation mode for high plasma densities > 1.5 1020 m-3, fulfilled the divertor conditions. This paper presents the approximate expressions for the optimal set of reactor parameters for rBS/I∼1, based on the self-consistent plasma simulations by 1.5D ASTRA code. The linear MHD stability analysis for ideal n=1 kink and ballooning modes has been carried out to determine the conditions of stabilization for bootstrap steady state tokamak reactor BSSTR configurations. (author) 10 refs., 1 tab
Siple Dome: Is it in Steady State?
Pettit, E. C.; Waddington, E. D.; Nereson, N. A.; Zumberge, M. A.; Hamilton, G. S.
2001-12-01
Changes in the West Antarctic Ice Sheet since the end of the last ice age have implications for how we interpret its present behavior, in terms of both its stability and its record of climate history. Siple Dome, the ridge between Ice Streams C and D, is not presently thinning and is close to being in balance with present environmental conditions. We present three independent measurements of ice thickness change in the divide region of Siple Dome: a GPS surface horizontal strain network, fiber optic vertical strain measurements at depth, and precision GPS measurements of vertical motion of near-surface ice ("coffee-can" method). From the horizontal strain network, we calculate the divergence of the horizontal velocity. This divergence is equal to the gradient of vertical velocity at the surface and, with some assumptions about the distribution of strain rates with depth, we can calculate the vertical velocity at the surface. For steady state, the vertical velocity must be balanced by the local accumulation rate. The fiber optic instruments provide a profile of the relative vertical velocity with depth. We fit a theoretical vertical velocity pattern to these data and extrapolate to find the surface vertical velocity. Our third method (coffee-can) directly measures the vertical motion of a marker 20 meters deep using precision GPS and compares it with the local long-term rate of snow accumulation to calculate the net rate of ice sheet thickness change. All three methods reach the same conclusion: Siple Dome is currently very close to being in steady state. This result has two implications. First, ice dynamics models developed to interpret radar images or ice core data can assume steady state behavior, simplifying the models. Second, our result suggests that the central part of the Ross Embayment may have had a low-elevation profile during the late Holocene, even though other areas of the WAIS may have been thicker.
Steady state plasma operation in RF dominated regimes on EAST
Zhang, X. J.; Zhao, Y. P.; Gong, X. Z.; Hu, C. D.; Liu, F. K.; Hu, L. Q.; Wan, B. N., E-mail: bnwan@ipp.ac.cn; Li, J. G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)
2015-12-10
Significant progress has recently been made on EAST in the 2014 campaign, including the enhanced CW H&CD system over 20MW heating power (LHCD, ICRH and NBI), more than 70 diagnostics, ITER-like W-monoblock on upper divertor, two inner cryo-pumps and RMP coils, enabling EAST to investigate long pulse H mode operation with dominant electron heating and low torque to address the critical issues for ITER. H-mode plasmas were achieved by new H&CD system or 4.6GHz LHCD alone for the first time. Long pulse high performance H mode has been obtained by LHCD alone up to 28s at H{sub 98}∼1.2 or by combing of ICRH and LHCD, no or small ELM was found in RF plasmas, which is essential for steady state operation in the future Tokamak. Plasma operation in low collision regimes were implemented by new 4.6GHz LHCD with core Te∼4.5keV. The non-inductive scenarios with high performance at high bootstrap current fraction have been demonstrated in RF dominated regimes for long pulse operation. Near full non-inductive CD discharges have been achieved. In addition, effective heating and decoupling method under multi-transmitter for ICRF system were developed in this campaign, etc. EAST could be in operation with over 30MW CW heating and current drive power (LHCD ICRH NBI and ECRH), enhanced diagnostic capabilities and full actively-cooled metal wall from 2015. It will therefore allow to access new confinement regimes and to extend these regimes towards to steady state operation.
Characteristics of edge-localized modes in the experimental advanced superconducting tokamak (EAST)
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....
ITER steady-state magnetic sensors: design status and performance
(private communication); [2] P. Moreau et al, Fusion Engineering and Design 84 (2009) 1344-1350; [3] ITER Design Description Document 55.A5,A6 Outer Vessel Steady State Sensors (private communication); [4] GRT047 Technical Report on Equilibrium reconstruction in ITER using external pick-up and steady state sensors, in preparation (private communication); [5] GRT047 Technical Report on Total toroidal current reconstruction in ITER using external pick-up and steady state sensors, in preparation (private communication). (authors)
Steady state phreatic surfaces in sloping aquifers
Loáiciga, Hugo A.
2005-08-01
Steady state groundwater flow driven by constant recharge in an unconfined aquifer overlying sloping bedrock is shown to be represented, using the Dupuit approximation, by an ordinary differential equation of the Abel type y(x) . y'(x) + a . y(x) + x = 0, whose analytical solution is derived in this work. This article first investigates the case of zero saturated thickness at the upstream boundary, a flow system reminiscent of perched groundwater created by percolation of precipitation or irrigation in a sloping aquifer fully draining at its downstream boundary. A variant of this flow system occurs when the phreatic surface mounds and produces groundwater discharge toward the upstream boundary. This variant is a generalization of the classical groundwater flow problem involving two lakes connected by an aquifer, the latter being on sloping terrain in this instance. Analytical solutions for the phreatic surface's steady state geometry are derived for the case of monotonically declining hydraulic head as well as for the case of a mounded phreatic surface. These solutions are of practical interest in drainage studies, slope stability, and runoff formation investigations. It is shown that the flow factor a = -$\\sqrt{{\\rm K}/{\\rm N} tan β (where K, N, and tan β are the hydraulic conductivity, vertical recharge, and aquifer slope, respectively) has a commanding role on the phreatic surface's solutions. Two computational examples illustrate the implementation of this article's results.
Advanced ICRF antenna design for R-TOKAMAK
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)
Magnetic confinement experiment. I: Tokamaks
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
Steady states of the parametric rotator and pendulum
Bouzas, Antonio O, E-mail: abouzas@fis.mda.cinvestav.m [Departamento de Fisica Aplicada, CINVESTAV-IPN, Carretera Antigua a Progreso Km. 6, Apdo Postal 73 ' Cordemex' , Merida 97310, Yucatan (Mexico)
2010-11-15
We discuss several steady-state rotation and oscillation modes of the planar parametric rotator and pendulum with damping. We consider a general elliptic trajectory of the suspension point for both rotator and pendulum, for the latter at an arbitrary angle with gravity, with linear and circular trajectories as particular cases. We treat the damped, nonlinear equation of motion of the parametric rotator and pendulum perturbatively for small parametric excitation and damping, although our perturbative approach can be extended to other regimes as well. Our treatment involves only ordinary second-order differential equations with constant coefficients, and provides numerically accurate perturbative solutions in terms of elementary functions. Some of the steady-state rotation and oscillation modes studied here have not been discussed in the previous literature. Other well-known ones, such as parametric resonance and the inverted pendulum, are extended to elliptic parametric excitation tilted with respect to gravity. The results presented here should be accessible to advanced undergraduates, and of interest to graduate students and specialists in the field of nonlinear mechanics.
Steady states of the parametric rotator and pendulum
We discuss several steady-state rotation and oscillation modes of the planar parametric rotator and pendulum with damping. We consider a general elliptic trajectory of the suspension point for both rotator and pendulum, for the latter at an arbitrary angle with gravity, with linear and circular trajectories as particular cases. We treat the damped, nonlinear equation of motion of the parametric rotator and pendulum perturbatively for small parametric excitation and damping, although our perturbative approach can be extended to other regimes as well. Our treatment involves only ordinary second-order differential equations with constant coefficients, and provides numerically accurate perturbative solutions in terms of elementary functions. Some of the steady-state rotation and oscillation modes studied here have not been discussed in the previous literature. Other well-known ones, such as parametric resonance and the inverted pendulum, are extended to elliptic parametric excitation tilted with respect to gravity. The results presented here should be accessible to advanced undergraduates, and of interest to graduate students and specialists in the field of nonlinear mechanics.
Technical Challenges in the Construction of the Steady-State Stellarator Wendelsetin 7-X
Full text: The 'fully-optimized' stellarator Wendelstein 7-X stellarator, presently under construction in Greifswald, combines a quasi-isodynamic magnetic field configuration sustained by superconducting coils with a steady-state exhaust concept, steady-state heating at high power, and a size sufficient to reach reactor-relevant nΤτ-values. It is the mission of the project to demonstrate the reactor potential of the optimized stellarator line. For the development of a credible stellarator reactor concept, steady-state operation has to be demonstrated with fully integrated discharge scenarios at high heating power with a divertor providing suitable power and particle exhaust. The development of reactor-relevant operation regimes is the chief scientific goal of Wendelstein 7-X. The subject of steady-state operation, however, is of more general interest, as this is also of great concern and interest for future tokamak devices. Consistent with the physics requirements of steady-state plasmas must be the engineering aspects of a steady-state fusion device. We discuss these issues for the design, manufacturing, and assembly of Wendelstein 7-X. The major components of Wendelstein 7-X have been manufactured, tested and delivered: 70 super-conducting coils, 121 superconducting bus-bars for the 7 coil current circuits, about 1000 cryo pipes, 10 half-modules of the central support structure, the plasma vessel and outer vessel, and 254 ports. The main focus of the project has in recent years shifted to the assembly process and considerable progress has been achieved. Although in the early phases of the Wendelstein 7-X construction several schedule delays have accumulated, there have been no major project delays for more than four years and completion of the device is foreseen for mid 2014. A summary of the technological challenges that have been faced in the project and solutions found are discussed in this paper. In addition the route towards completion, commissioning, and
The ARIES Advanced and Conservative Tokamak Power Plant Study
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
Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches
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
Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches
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.
Influence of helium puff on divertor asymmetry in experimental advanced superconducting tokamak
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...
Steady-state creep in the mantle
G. RANALLI
1977-06-01
Full Text Available SUMMARY - The creep equations for steady-state flow of olivine at high
pressure and temperature are compared in an attempt to elucidate the rheological
behaviour of the mantle. Results are presented in terms of applied deformation
maps and curves of effective viscosity v depth.
In the upper mantle, the transition stress between dislocation and diffusion
creep is between 10 to 102 bar (as orders of magnitude for grain sizes from
0.01 to 1 cm. The asthenosphere under continents is deeper, and has higher
viscosity, than under oceans. Predominance of one creep mechanism above the
others depends on grain size, strain rate, and volume fraction of melt; the
rheological response can be different for different geodynamic processes.
In the lower mantle, on the other hand, dislocation creep is predominant
at all realistic grain sizes and strain rates. If the effective viscosity has to be only
slightly higher than in the upper mantle, as some interpretations of glacioisostatic
rebound suggest, then the activation volume cannot be larger than
11 cm3 mole^1.
ARIES-AT: An advanced tokamak, advanced technology fusion power plant
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)
ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM
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.
Constrained optimal steady-state control for isolated traffic intersections
Jack HADDAD; David MAHALEL; Ilya IOSLOVICH; Per-Olof GUTMAN
2014-01-01
The steady-state or cyclic control problem for a simplified isolated traffic intersection is considered. The optimization problem for the green-red switching sequence is formulated with the help of a discrete-event max-plus model. Two steady-state control problems are formulated: optimal steady-state with green duration constraints, and optimal steady-state control with lost time. In the case when the criterion is a strictly increasing, linear function of the queue lengths, the steady-state control problems can be solved analytically. The structure of constrained optimal steady-state traffic control is revealed, and the effect of the lost time on the optimal solution is illustrated.
Steady-State Chemotactic Response in E. coli
Kafri, Yariv
2007-01-01
The bacterium E. coli maneuvers itself to regions with high chemoattractant concentrations by performing two stereotypical moves: `runs', in which it moves in near straight lines, and `tumbles', in which it does not advance but changes direction randomly. The duration of each move is stochastic and depends upon the chemoattractant concentration experienced in the recent past. We relate this stochastic behavior to the steady-state density of a bacterium population, and we derive the latter as a function of chemoattractant concentration. In contrast to earlier treatments, here we account for the effects of temporal correlations and variable tumbling durations. A range of behaviors obtains, that depends subtly upon several aspects of the system - memory, correlation, and tumbling stochasticity in particular.
An Adsorption Equilibria Model for Steady State Analysis
Ismail, Azhar Bin
2016-02-29
The investigation of adsorption isotherms is a prime factor in the ongoing development of adsorption cycles for a spectrum of advanced, thermally-driven engineering applications, including refrigeration, natural gas storage, and desalination processes. In this work, a novel semi-empirical mathematical model has been derived that significantly enhances the prediction of the steady state uptake in adsorbent surfaces. This model, a combination of classical Langmuir and a novel modern adsorption isotherm equation, allows for a higher degree of regression of both energetically homogenous and heterogeneous adsorbent surfaces compared to several isolated classical and modern isotherm models, and has the ability to regress isotherms for all six types under the IUPAC classification. Using a unified thermodynamic framework, a single asymmetrical energy distribution function (EDF) has also been proposed that directly relates the mathematical model to the adsorption isotherm types. This fits well with the statistical rate theory approach and offers mechanistic insights into adsorption isotherms.
Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak.
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
Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak
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
Development of a plasma control system for steady-state operation on QUEST
A drift error correction technique with machine vision and a real-time equilibrium calculation code have been developed on the QUEST (Q-shu university experiment with the steady-state spherical tokamak) for steady-state operation. The drift error caused by the long time-integration of magnetic raw signals has to be removed. With a captured image of the plasma's cross section, the plasma's position is identified by use of image filters. The measured magnetic flux values are corrected to the calculated flux values estimated by using this plasma position. The correction with the captured image work as expected in the preliminary result using a flashlight instead of a plasma.
Development of a plasma control system for steady-state operation on QUEST
Hasegwa, Makoto; Nakamura, Kazuo; Zushi, Hideki [Kyushu University, Fukuoka (Japan); and others
2014-10-15
A drift error correction technique with machine vision and a real-time equilibrium calculation code have been developed on the QUEST (Q-shu university experiment with the steady-state spherical tokamak) for steady-state operation. The drift error caused by the long time-integration of magnetic raw signals has to be removed. With a captured image of the plasma's cross section, the plasma's position is identified by use of image filters. The measured magnetic flux values are corrected to the calculated flux values estimated by using this plasma position. The correction with the captured image work as expected in the preliminary result using a flashlight instead of a plasma.
Development of a plasma control system for steady-state operation on QUEST
Hasegwa, Makoto; Nakamura, Kazuo; Zushi, Hideki; Hanada, Kazuaki; Fujisawa, Akihide; Matsuoka, Keisuke; Idei, Hiroshi; Nagashima, Yoshihiko; Tokunaga, Kazutoshi; Kawasaki, Shoji; Nakashima, Hisatoshi; Higashijima, Aki
2014-10-01
A drift error correction technique with machine vision and a real-time equilibrium calculation code have been developed on the QUEST (Q-shu university experiment with the steady-state spherical tokamak) for steady-state operation. The drift error caused by the long time-integration of magnetic raw signals has to be removed. With a captured image of the plasma's cross section, the plasma's position is identified by use of image filters. The measured magnetic flux values are corrected to the calculated flux values estimated by using this plasma position. The correction with the captured image work as expected in the preliminary result using a flashlight instead of a plasma.
Data acquisition and control system for steady state neutral beam injector
This paper presents the control system overview, hardware, software and network for Data acquisition and Control system for steady state neutral beam injector (NBIDACS) to be used for heating of plasma in steady state superconducting tokamak (SST-1). The task for NBIDACS is not only to safely deliver 1.7 MW of neutral beams at 55 keV H deg. a period of 1000 s with 16.7% duty cycle but also to acquire the data related to house keeping of the system and its auxiliaries and diagnostics which determine the quality and parameters of the beam. Major issues concerning the design of the system stem from operation duty cycle of 1000 s ON/5000 s OFF. This calls for use of intelligent techniques not only for managing a large amount (100 MB) of data per shot but also to obtain failsafe, reliable control system and to archive the recorded data
Comparison between a steady-state fusion reactor and an inductively driven pulse reactor
In the present report, a comparison is made between tokamak reactors of steady state operation -SSTR- and pulse operation. The former design uses neutral beams as a current driver to realize steady state operation. The latter is inductively operated basic tokamak with burn time of one hour to a half day. This time is determined by dimensions of the central solenoid coil and these dimensions also determine the basic design concept of the pulse tokamak. The dimension includes effect of fatigue due to pulse operation. Performance as a power plant is evaluated with a schematic design of heat transport and power generation system. Heat accumulation in the primary coolant loop is studied in order to make up for a dwell time of a pulse reactor. It is shown that large heat accumulator is necessary to suppress a drop in output during the dwell time. The dwell time has an optimum length with respect to the dwell time. Comparison of fusion plant with other energy source reveals that reduction of the size is essential in order that the fusion is competitive with other sources. (author)
Defining Features of Steady-State Timbres
Hall, Michael D.
1995-01-01
Three experiments were conducted to define steady -state features of timbre for a group of well-trained musicians. Experiment 1 evaluated whether or not pairs of three critical dimensions of timbre--spectral slope (6 or 12 dB/octave), formant structure (/a/ or /i/ vowel), and inharmonicity of partials (harmonic or inharmonic)--were processed in a separable or integral fashion. Accuracy and speed for classification of values along one dimension were examined under different conditions of variability along a second dimension (fixed, correlated, or orthogonal). Spectral slope and formant structure were integral, with classification speed for the target dimension depending upon variability along the orthogonal dimension. In contrast, evidence of asymmetric separability was obtained for inharmonicity. Classification speed for slope and formant structure did not depend on inharmonicity, whereas RT for the target dimension of inharmonicity was strongly influenced by variability along either slope or formant structure. Since the results of Experiment 1 provided a basis for manipulating spectral slope and formant structure as a single feature, these dimensions were correlated in Experiment 2. Subjects searched for targets containing potential features of timbre within arrays of 1-4 inharmonic distractor pitches. Distractors were homogeneous with respect to the dimensions of timbre. When targets had /a/ formants with shallow spectral slopes, search time increased nonlinearly with array size in a manner consistent with the parallel processing of items, and thus feature search. Feature search was not obtained for targets with /i/ formants and steep slopes. Thus, the feature was coded as the presence or absence of /a/ formants with shallow spectral slopes. A search task using heterogeneous distractor values along slope/formant structure was used in Experiment 3 to evaluate whether or not the feature of timbre and pitch were automatically conjoined (integral). Search times for
Positive Steady States of a Competitor-Competitor-Mutualist Model
Wen-yan Chen; Ming-xin Wang
2004-01-01
In this paper we deal with the positive steady states of a Competitor-Competitor-Mutualist model with diffusion and homogeneous Dirichlet boundary conditions.We rst give the necessary conditions,and then establish the su cient conditions for the existence of positive steady states.
INTEGATED ADVANCED TOKAMAK OPERATION ON DIII-D
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
Steady State of Pedestrian Flow in Bottleneck Experiments
Liao, Weichen; Seyfried, Armin; Chraibi, Mohcine; Drzycimski, Kevin; Zheng, Xiaoping; Zhao, Ying
2015-01-01
Experiments with pedestrians could depend strongly on initial conditions. Comparisons of the results of such experiments require to distinguish carefully between transient state and steady state. In this work, a feasible algorithm - Cumulative Sum Control Chart - is proposed and improved to automatically detect steady states from density and speed time series of bottleneck experiments. The threshold of the detection parameter in the algorithm is calibrated using an autoregressive model. Comparing the detected steady states with previous manually selected ones, the modified algorithm gives more reproducible results. For the applications, three groups of bottleneck experiments are analysed and the steady states are detected. The study about pedestrian flow shows that the difference between the flows in all states and in steady state mainly depends on the ratio of pedestrian number to bottleneck width. When the ratio is higher than a critical value (approximately 115 persons/m), the flow in all states is almost ...
Steady-State Performance of Kalman Filter for DPLL
QIAN Yi; CUI Xiaowei; LU Mingquan; FENG Zhenming
2009-01-01
For certain system models, the structure of the Kalman filter is equivalent to a second-order vari-able gain digital phase-locked loop (DPLL). To apply the knowledge of DPLLs to the design of Kalman filters, this paper studies the steady-state performance of Kalman filters for these system models. The results show that the steady-state Kalman gain has the same form as the DPLL gain. An approximate simple form for the steady-state Kalman gain is used to derive an expression for the equivalent loop bandwidth of the Kalman filter as a function of the process and observation noise variances. These results can be used to analyze the steady-state performance of a Kalman filter with DPLL theory or to design a Kalman filter model with the same steady-state performance as a given DPLL.
Fast-ion transport in qmin>2, high-β steady-state scenarios on DIII-D
Results from experiments on DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] aimed at developing high β steady-state operating scenarios with high-qmin confirm that fast-ion transport is a critical issue for advanced tokamak development using neutral beam injection current drive. In DIII-D, greater than 11 MW of neutral beam heating power is applied with the intent of maximizing βN and the noninductive current drive. However, in scenarios with qmin>2 that target the typical range of q95= 5–7 used in next-step steady-state reactor models, Alfvén eigenmodes cause greater fast-ion transport than classical models predict. This enhanced transport reduces the absorbed neutral beam heating power and current drive and limits the achievable βN. In contrast, similar plasmas except with qmin just above 1 have approximately classical fast-ion transport. Experiments that take qmin>3 plasmas to higher βP with q95= 11–12 for testing long pulse operation exhibit regimes of better than expected thermal confinement. Compared to the standard high-qmin scenario, the high βP cases have shorter slowing-down time and lower ∇βfast, and this reduces the drive for Alfvénic modes, yielding nearly classical fast-ion transport, high values of normalized confinement, βN, and noninductive current fraction. These results suggest DIII-D might obtain better performance in lower-q95, high-qmin plasmas using broader neutral beam heating profiles and increased direct electron heating power to lower the drive for Alfvén eigenmodes
UNIVERSAL THEORY OF STEADY-STATE ONE-DIMENSIONAL PHOTOREFRACTIVE SOLITONS
刘劲松
2001-01-01
A universal theory of steady-state one-dimensional photorefractive spatial solitons is developed which applies to the steady-state one-dimensional photorefractive solitons under various realizations, including the screening solitons in a biased photorefractive medium, the photovoltaic solitons in open- and closed-circuit photovoltaic-photorefractive media and the screening-photovoltaic solitons in biased photovoltaic-photorefractive media. Previous theories advanced individually elsewhere for these solitons can be obtained by simplifying the universal theory under the appropriate conditions.
ASPECT: An advanced specified-profile evaluation code for tokamaks
Stotler, D.P.; Reiersen, W.T.; Bateman, G.
1993-03-01
A specified-profile, global analysis code has been developed to evaluate the performance of fusion reactor designs. Both steady-state and time-dependent calculations are carried out; the results of the former can be used in defining the parameters of the latter, if desired. In the steady-state analysis, the performance is computed at a density and temperature chosen to be consistent with input limits (e.g., density and beta) of several varieties. The calculation can be made at either the intersection of the two limits or at the point of optimum performance as the density and temperature are varied along the limiting boundaries. Two measures of performance are available for this purpose: the ignition margin or the confinement level required to achieve a prescribed ignition margin. The time-dependent calculation can be configured to yield either the evolution of plasma energy as a function of time or, via an iteration scheme, the amount of auxiliary power required to achieve a desired final plasma energy.
ASPECT: An advanced specified-profile evaluation code for tokamaks
A specified-profile, global analysis code has been developed to evaluate the performance of fusion reactor designs. Both steady-state and time-dependent calculations are carried out; the results of the former can be used in defining the parameters of the latter, if desired. In the steady-state analysis, the performance is computed at a density and temperature chosen to be consistent with input limits (e.g., density and beta) of several varieties. The calculation can be made at either the intersection of the two limits or at the point of optimum performance as the density and temperature are varied along the limiting boundaries. Two measures of performance are available for this purpose: the ignition margin or the confinement level required to achieve a prescribed ignition margin. The time-dependent calculation can be configured to yield either the evolution of plasma energy as a function of time or, via an iteration scheme, the amount of auxiliary power required to achieve a desired final plasma energy
Measurement of non-steady-state free fatty acid turnover
The accuracy of non-steady-state equations for measuring changes in free fatty acid rate of appearance (Ra) is unknown. In the present study, endogenous lipolysis (traced with [14C]-linoleate) was pharmacologically suppressed in six conscious mongrel dogs. A computer-responsive infusion pump was then used to deliver an intravenous oleic acid emulsion in both constant and linear gradient infusion modes. Both non-steady-state equations with various effective volumes of distribution (V) and steady-state equations were used to measure oleate Ra [(14C]oleate). Endogenous lipolysis did not change during the experiment. When oleate Ra increased in a linear gradient fashion, only non-steady-state equations with a large (150 ml/kg) V resulted in erroneous values (9% overestimate, P less than 0.05). In contrast, when oleate Ra decreased in a similar fashion, steady-state and standard non-steady-state equations (V = plasma volume = 50 ml/kg) overestimated total oleate Ra (18 and 7%, P less than 0.001 and P less than 0.05, respectively). Overall, non-steady-state equations with an effective V of 90 ml/kg (1.8 x plasma volume) allowed the most accurate estimates of oleate Ra
Measurement of non-steady-state free fatty acid turnover
Jensen, M.D.; Heiling, V.; Miles, J.M. (Mayo Clinic, Rochester, MN (USA))
1990-01-01
The accuracy of non-steady-state equations for measuring changes in free fatty acid rate of appearance (Ra) is unknown. In the present study, endogenous lipolysis (traced with ({sup 14}C)-linoleate) was pharmacologically suppressed in six conscious mongrel dogs. A computer-responsive infusion pump was then used to deliver an intravenous oleic acid emulsion in both constant and linear gradient infusion modes. Both non-steady-state equations with various effective volumes of distribution (V) and steady-state equations were used to measure oleate Ra (({sup 14}C)oleate). Endogenous lipolysis did not change during the experiment. When oleate Ra increased in a linear gradient fashion, only non-steady-state equations with a large (150 ml/kg) V resulted in erroneous values (9% overestimate, P less than 0.05). In contrast, when oleate Ra decreased in a similar fashion, steady-state and standard non-steady-state equations (V = plasma volume = 50 ml/kg) overestimated total oleate Ra (18 and 7%, P less than 0.001 and P less than 0.05, respectively). Overall, non-steady-state equations with an effective V of 90 ml/kg (1.8 x plasma volume) allowed the most accurate estimates of oleate Ra.
Development on JET of Advanced Tokamak Operations for ITER
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)
A Note on Equations for Steady-State Optimal Landscapes
Liu, H.H.
2010-06-15
Based on the optimality principle (that the global energy expenditure rate is at its minimum for a given landscape under steady state conditions) and calculus of variations, we have derived a group of partial differential equations for describing steady-state optimal landscapes without explicitly distinguishing between hillslopes and channel networks. Other than building on the well-established Mining's equation, this work does not rely on any empirical relationships (such as those relating hydraulic parameters to local slopes). Using additional constraints, we also theoretically demonstrate that steady-state water depth is a power function of local slope, which is consistent with field data.
The relevant parameters of two steady-state models of a plasma column, in fusion regime, were analyzed for an ideal Tokamak. The neo-classical transport theory was considered in the banana regime and in the Pfirsch-Schlueter regime. The first model proposes a correction in the numerical coefficients of the transport equations. In the other one, a poloidal current from Pfirsch-Schlueter classical diffusion is considered aiming to satisfy the pressure balance. (M.C.K.)
Enhancement of the steady-state magnetization in TROSY experiments
Under the condition that the longitudinal relaxation time of spin I is shorter than the longitudinal relaxation time of spin S the steady-state magnetization in [S,I]-TROSY-type experiments can be enhanced by intermediate storage of a part of the steady-state magnetization of spin I on spin S with a pulse sequence element during the relaxation delay. It is demonstrated with samples ranging in size from the 1 kDa cyclosporin to the 110 kDa 15N,2H-labeled dihydroneopterin Aldolase that intermediate storage of steady-state magnetization in a [15N,1H]-TROSY experiment yields a signal gain of 10-25%. The method proposed here for intermediate storage of steady-state magnetization can be implemented in any [15N,1H]-TROSY-type experiments
First experiments with SST-1 tokamak
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
Steady-state current transfer and scattering theory
Ben-Moshe, Vered; Rai, Dhurba; Skourtis, Spiros S.; Nitzan, Abraham
2010-01-01
The correspondence between the steady state theory of current transfer and scattering theory in a system of coupled tight-binding models of 1-dimensional wires is explored. For weak interwire coupling both calculations give nearly identical results, except at singular points associated with band edges. The effect of decoherence in each of these models is studied using a generalization of the Liouville-von Neuman equation suitable for steady-state situations. An example of a single impurity mo...
Steady-state leaching of tritiated water from silica gel
Das, H.A.; Hou, Xiaolin
2009-01-01
Aqueous leaching of tritium from silica gel, loaded by absorption of water vapor, makes part of reactor de-commissioning. It is found to follow the formulation of steady-state diffusion.......Aqueous leaching of tritium from silica gel, loaded by absorption of water vapor, makes part of reactor de-commissioning. It is found to follow the formulation of steady-state diffusion....
TRANSIENT AND STEADY-STATE DYNAMICS OF GRANULAR SHEAR FLOWS
Losert, W.; Kwon, G.
2001-01-01
The initiation and steady-state dynamics of granular shear flow are investigated experimentally in a Couette geometry with independently moveable outer and inner cylinders. The motion of particles on the top surface is analyzed using fast imaging. During steady state rotation of both cylinders at different rates, a shear band develops close to the inner cylinder for all combinations of speeds of each cylinder we investigated. Experiments on flow initiation were carried out with one of the cyl...
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
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.
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
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
The impact of ripple on the design of advanced fuel tokamaks
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
The ARIES-AT advanced tokamak, Advanced technology fusion power plant
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
Numerical study of Alfvén eigenmodes in the Experimental Advanced Superconducting Tokamak
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
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...
The contribution to the energy balance and transport in an advanced-fuel tokamak reactor
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
Lake contamination models for evolution towards steady state
Johan C. VAREKAMP
2003-09-01
Full Text Available Most lakes are in an average steady state for water but contaminants may not yet have reached steady state or are gradually being flushed out in a clean-up program. The evolution towards steady state for fully mixed or stratified lakes can be described by basic equations of mass flow. The time-concentration paths for fully mixed lakes are asymptotic toward a steady state concentration, which is reached in about 6 contaminant residence times (and clean-up also takes about 6 residence times. Stratified lakes also evolve towards a whole-lake steady state concentration but show oscillating patterns of concentration versus time, with the amplitude and dampening period depending on the volume ratio of epilimnion to total lake volume. In most natural lakes, the compositional contrast between epilimnion and hypolimnion will become almost erased in 2-4 residence times. An acid lake in North-Patagonia is used as an example of contamination of a thermally stratified lake by volcanic effluents.
Nonequilibrium steady states in fluids of platelike colloidal particles
Bier, Markus; van Roij, René
2008-02-01
Nonequilibrium steady states in an open system connecting two reservoirs of platelike colloidal particles are investigated by means of a recently proposed phenomenological dynamic density functional theory [M. Bier and R. van Roij, Phys. Rev. E 76, 021405 (2007)]. The platelike colloidal particles are approximated within the Zwanzig model of restricted orientations, which exhibits an isotropic-nematic bulk phase transition. Inhomogeneities of the local chemical potential generate a diffusion current which relaxes to a nonvanishing value if the two reservoirs coupled to the system sustain different chemical potentials. The relaxation process of initial states towards the steady state turns out to comprise two regimes: a smoothening of initial steplike structures followed by an ultimate relaxation of the slowest diffusive mode. The position of a nonequilibrium interface and the particle current of steady states depend nontrivially on the structure of the reservoirs due to the coupling between translational and orientational degrees of freedom of the fluid.
Stable MIMO Constrained Predictive Control with Steady state Objective Optimization
无
2000-01-01
A two-stage multi-objective optimization model-predictive control algorithms(MPC) strategy is pre sented. A domain MPC controller with input constraints is used to increase freedom for steady-state objective and enhance stabilization of the controller. A steady-state objective optimization algorithm oriented to transient process is adopted to realize optimization of objectives else than dynamic control. It is proved that .the stabilization for both dynamic control and steady-state objective optimization can be guaranteed. The theoretical results are demonstrated and discussed using a distillation tower as the model. Theoretical analysis and simulation results show that this control strategy is efficient and provides a good strategic solution to practical process control.
From Steady-State To Cyclic Metal Forming Processes
Montmitonnet, Pierre
2007-05-01
Continuous processes often exhibit a high proportion of steady state, and have been modeled with steady-state formulations for thirty years, resulting in very CPU-time efficient computations. On the other hand, incremental forming processes generally remain a challenge for FEM software, because of the local nature of deformation compared with the size of the part to be formed, and of the large number of deformation steps needed. Among them however, certain semi-continuous metal forming processes can be characterized as periodic, or cyclic. In this case, an efficient computational strategy can be derived from the ideas behind the steady-state models. This will be illustrated with the example of pilgering, a seamless tube cold rolling process.
Avoiding Rebound through a Steady-State Economy
Nørgaard, Jørgen
conditions in many parts of the world, the transition towards a steady-state economy needs to begin first in the affluent countries, including the Nordic countries from where most of the information in this chapter is drawn. The politicians in these countries are not seeking a steady-state economy, but some...... only buy some time. From this perspective, the environmental problem with the rebound effect is not the higher energy efficiency, which pushes towards lower flows of resources through the economy, but rather the conventional economy which rebounds the savings, because of its quest for higher flows. In...... this chapter, I shall take the rebound debate further by discussing the possible role of energy efficiency in a sustainable economy that is based on the notion of ‘sufficiency’. The assumption is that globally we need to achieve a ‘steady-state economy’. Considering the urgent need for better material...
Free Boundary Problem of Ono—steady State Seepage Flow
XiaomingGUO; Ying－SUN; 等
1999-01-01
Along with the vigorous developing construction,the number of various underground engineerings is greatly increasing,Such as:the foundations of dams and high-rise multistoried houses,subways and tunnels,water and oil wells etc., where the close attention is always payed to the seepage behaviour in the media around the strutures.The Variatonal Inequality formulation and its FEM solution for the free boundary problem of 2D steady state seepage flow was given by the authors,In this paper a further investigation is made on the non-steady state seepage problem,taken the seepage flow of wells as an example.The presented approach-Variational Inequality and its FEM solution-is also very beneficial to the non-steady state problems,where the transient free boundary can also be defined directly without conventional iterations.
Thermalization of Starlight in the Steady-State Cosmology
Ibison, M
2009-01-01
We investigate the fate of starlight in the Steady-State Cosmology. We discover that it is largely unaffected by the presence of ions in intergalactic space as it gets progressively red-shifted from the visible all the way down to the plasma frequency of the intergalactic matter. At that point, after about 450 Gyr - and contrary to previously published claims - the radiation will be thermalized. Under the assumptions adopted by Gold, Bondi, Hoyle, Narlikar, Burbidge and others concerning the creation of matter in the Steady-State Cosmology, and using reasonable estimates for the baryonic mass-density and mass-fraction of 4He, the analysis predicts a universal radiation field matching the CMB, i.e. having a black-body spectrum and temperature of about 2.7 K. The Steady-state Cosmology predicts that this radiation field will appear to originate from the intergalactic plasma.
Steady-state and transient wellbore temperatures during drilling
McDonald, W.J.
1976-05-20
An extensive literature search was made to locate technical publications and computer programs relating to wellbore temperatures during drilling operations. The search confirmed the need for knowledge of transient and steady state circulating temperatures in the design of geothermal bits. Two approaches were used in calculating borehole temperatures. The steady state solution of Holmes and Swift was programmed and 2100 cases calculated for various borehole configurations. For transient temperature studies, calculations were made for ten borehole configurations. These calculations help emphasize the need for better high temperature bit performance and improved engineering procedures in drilling. The conclusions and recommendations are based on latest available technology for calculating wellbore temperatures.
Correlation Between Steady State and Impulse Earth Resistance Values
N. M. Nor
2009-01-01
Full Text Available This study presented experimental results of earthing systems under low-magnitude currents and under high impulse currents. The details of the measuring circuit involved for both types of testing were described. Three field sites were selected. At each site, three earth electrodes configurations were used. This makes up to nine earthing systems. From both low magnitude and impulse tests, the correlation between the steady state earth resistance value and the earth resistance under fast impulse currents can be observed. The relation between the calculated and measured steady state earth resistance is also shown in this study.
Steady-state entanglement activation in optomechanical cavities
Farace, Alessandro; Ciccarello, Francesco; Fazio, Rosario; Giovannetti, Vittorio
2014-02-01
Quantum discord, and related indicators, are raising a relentless interest as a novel paradigm of nonclassical correlations beyond entanglement. Here, we discover a discord-activated mechanism yielding steady-state entanglement production in a realistic continuous-variable setup. This comprises two coupled optomechanical cavities, where the optical modes (OMs) communicate through a fiber. We first use a simplified model to highlight the creation of steady-state discord between the OMs. We show next that such discord improves the level of stationary optomechanical entanglement attainable in the system, making it more robust against temperature and thermal noise.
Electric machines steady state, transients, and design with Matlab
Boldea, Ion
2009-01-01
Part I: Steady StateIntroductionElectric Energy and Electric MachinesBasic Types of Transformers and Electric MachinesLosses and EfficiencyPhysical Limitations and RatingsNameplate RatingsMethods of AnalysisState of the Art and Perspective Electric TransformersAC Coil with Magnetic Core and Transformer Principles Magnetic Materials in EMs and Their LossesElectric Conductors and Their Skin EffectsComponents of Single- and 3-Phase TransformersFlux Linkages and Inductances of Single-Phase TransformersCircuit Equations of Single-Phase Transformers With Core LossesSteady State and Equivalent Circui
Influence of the epithermal effects on the MCF steady state
This work is devoted to the correct interpretation of the steady-state parameters of the muon catalyzed fusion (MCF) process in a D/T mixture. Previously the influence of the epithermal effects (dtμ-molecule formation by 'hot', non-thermalized tμ atoms) on the steady-state parameters was studied only for measurements with a low-density target (density φ=0.01 relative to the liquid hydrogen density). We suggest a new method allowing direct determination of the necessary corrections to the MCF cycling rate for high-density data (φ≥0.4)
OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM
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
On the steady states of weakly reversible chemical reaction networks
Deng, Jian; Jones, Christopher; Feinberg, Martin; Nachman, Adrian
2011-01-01
A natural condition on the structure of the underlying chemical reaction network, namely weak reversibility, is shown to guarantee the existence of an equilibrium (steady state) in each positive stoichiometric compatibility class for the associated mass-action system. Furthermore, an index formula is given for the set of equilibria in a given stoichiometric compatibility class.
Optimising performance in steady state for a supermarket refrigeration system
Green, Torben; Kinnaert, Michel; Razavi-Far, Roozbeh; Izadi-Zamanabadi, Roozbeh; Niemann, Hans Henrik
Using a supermarket refrigeration system as an illustrative example, the paper postulates that by appropriately utilising knowledge of plant operation, the plant wide performance can be optimised based on a small set of variables. Focusing on steady state operations, the total system performance is...
Plasticity, Fracture and Friction in Steady-State Plate Cutting
Simonsen, Bo Cerup; Wierzbicki, Tomasz
1997-01-01
A closed form solution to the problem of steady-state wedge cutting through a ductile metal plate is presented. The considered problem is an idealization of a ship bottom raking process, i.e. a continuous cutting damage of a ship bottom by a hard knife-like rock in a grounding event. A new...
Plasticity, Fracture and Friction in Steady-State Plate Cutting
Simonsen, Bo Cerup; Wierzbicki, Tomasz
1997-01-01
A closed form solution to the problem of steady-state wedge cutting through a ductile metal plate is presented. The considered problem is an idealization of a ship bottom raking process, i.e. a continuous cutting damage of a ship bottom by a hard knife-like rock in a grounding event. A new kinema...
SBWR Model for Steady-State and Transient Analysis
Gilberto Espinosa-Paredes; Alejandro Nuñez-Carrera
2008-01-01
This paper presents a model of a simplified boiling water reactor (SBWR) to analyze the steady-state and transient behavior. The SBWR model is based on approximations of lumped and distributed parameters to consider neutronics and natural circulation processes. The main components of the model are vessel dome, downcomer, lower plenum, core (ch...
Dark Entangled Steady States of Interacting Rydberg Atoms
Dasari, Durga; Mølmer, Klaus
2013-01-01
their short-lived excited states lead to rapid, dissipative formation of an entangled steady state. We show that for a wide range of physical parameters, this entangled state is formed on a time scale given by the strengths of coherent Raman and Rabi fields applied to the atoms, while it is only weakly...
A displacement based FE formulation for steady state problems
Yu, Yuhong
2005-01-01
In this thesis a new displacement based formulation is developed for elasto-plastic deformations in steady state problems. In this formulation the displacements are the primary variables, which is in contrast to the more common formulations in terms of the velocities as the primary variables. In a s
Steady-State Pharmacokinetics of Bupropion SR in Juvenile Patients
Daviss, W. Burleson; Perel, James M.; Rudolph, George R.; Axelson, David A.; Gilchrist, Richard; Nuss, Sharon; Birmaher, Boris; Brent, David A.
2005-01-01
Objective: To examine the steady-state pharmacokinetic properties of bupropion sustained release (SR) and their potential developmental differences in youths. Method: Eleven boys and eight girls aged 11 to 17 years old were prescribed bupropion SR monotherapy for attention-deficit/hyperactivity disorder (n = 16) and/or depressive disorders (n =…
Combined Steady-State and Dynamic Heat Exchanger Experiment
Luyben, William L.; Tuzla, Kemal; Bader, Paul N.
2009-01-01
This paper describes a heat-transfer experiment that combines steady-state analysis and dynamic control. A process-water stream is circulated through two tube-in-shell heat exchangers in series. In the first, the process water is heated by steam. In the second, it is cooled by cooling water. The equipment is pilot-plant size: heat-transfer areas…
Extending the quasi-steady state approximation by changing variables
Borghans, J.A.M.; Boer, R.J. de; Segel, L.A.
1996-01-01
The parameter domain for which the quasi-steady state assumption is valid can be considerably extended merely by a simple change of variable. This is demonstrated for a variety of biologically significant examples taken from enzyme kinetics, immunology and ecology.
Steady state nutrition by transpiration controlled nutrient supply
Braakhekke, W.G.; Labe, D.A.
1990-01-01
Programmed nutrient addition with a constant relative addition rate has been advocated as a suitable research technique for inducing steady state nutrition in exponentially growing plants. Transpiration controlled nutrient supply is proposed as an alternative technique for plants with a short or no
Analysis of steady-state hydraulic tests in fractured rock
A model for the analysis of steady-state hydraulic injection tests into single fractures of a rock-mass is presented, and solved analytically. It is used to obtain a probability distribution for the transmissivities of fractures in Cornish granite. (author)
Three novel tokamak plasma regimes in TFTR
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
Three novel tokamak plasma regimes in TFTR
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.
Profile control of advanced tokamak plasmas in view of continuous operation
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.
Profile control of advanced tokamak plasmas in view of continuous operation
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.
Profile control of advanced tokamak plasmas in view of continuous operation
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
Advances in comprehensive gyrokinetic simulations of transport in tokamaks
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)
Action-at-a-distance electrodynamics in quasi-steady-state cosmology
Kaustubh Sudhir Deshpande
2014-09-01
Action-at-a-distance electrodynamics – alternative approach to field theory – can be extended to cosmological models using conformal symmetry. An advantage of this is that, the origin of arrow of time in electromagnetism can be attributed to the cosmological structure. Different cosmological models can be investigated, based on Wheeler–Feynman absorber theory, and only those models can be considered viable for our Universe which have net full retarded electromagnetic interactions, i.e., forward direction of time. This work evaluates the quasi-steady-state model and demonstrates that it admits full retarded and not advanced solution. Thus, quasi-steady-state cosmology (QSSC) satisfies this necessary condition for a correct cosmological model, based on action-at-a-distance formulation.
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
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.
Advanced Tokamak Regimes in Alcator C-Mod with Lower Hybrid Current Drive
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.
Towards steady-state operational design for the data and PF control systems of the HT-7U
Fusion energy is an ultimate and inexhaustible source of energy for mankind and is expected to be obtained in controlled operation within this century. Among various possible candidates for fusion, the tokamak is presently the most qualified one, and since it uses superconducting magnetic coils, it will be adequate for steady-state operation. The HT-7U superconducting tokamak is a part of national project in China on fusion research, scheduled to become available on-line by the end of 2004 (Wan Y.X. and HT-7 and HT-7U Groups 2000 Overview of steady state operation of HT-7 and present status of the HT-7U project Nucl. Fusion 40 1057). The control system of the HT-7U is designed as a distributed control system (HT7UDCS), including many subsystems that provide the various functions of supervision, remote control, real-time monitoring, data acquisition and data handling. The major features of the HT-7U tokamak, which make long-pulse (∼1000 s) operation possible are the flexible poloidal field (PF) system, an auxiliary heating system, the current-driving system and a divertor system. In order to realize these features simultaneously, real-time data handling and analysis, along with a significant control capability is required. This paper discusses the design of the HT7UDCS. (author)
The Steady State Calculation for SMART with MIDAS/SMR
KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. Before the severe accident sequences are estimated, it is prerequisite that MIDAS code predicts the steady state conditions properly. But MIDAS code does not include the heat transfer model for the helical tube. Therefore, the heat transfer models for the helical tube from TASS/SMR-S were implemented into MIDAS code. To estimate the validity of the implemented heat transfer correlations for the helical tube and the input data, the steady state was recalculated with MIDAS/SMR based on design level 2 and compared with the design values
The Steady State Calculation for SMART with MIDAS/SMR
Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seong Won [KORTIC, Daejeon (Korea, Republic of)
2010-10-15
KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. Before the severe accident sequences are estimated, it is prerequisite that MIDAS code predicts the steady state conditions properly. But MIDAS code does not include the heat transfer model for the helical tube. Therefore, the heat transfer models for the helical tube from TASS/SMR-S were implemented into MIDAS code. To estimate the validity of the implemented heat transfer correlations for the helical tube and the input data, the steady state was recalculated with MIDAS/SMR based on design level 2 and compared with the design values
Nonequilibrium Steady States of a Stochastic Model System.
Zhang, Qiwei
We study the nonequilibrium steady state of a stochastic lattice gas model, originally proposed by Katz, Lebowitz and Spohn (Phys. Rev. B 28: 1655 (1983)). Firstly, we solve the model on some small lattices exactly in order to see the general dependence of the steady state upon different parameters of the model. Nextly, we derive some analytical results for infinite lattice systems by taking some suitable limits. We then present some renormalization group results for the continuum version of the model via field theoretical techniques, the supersymmetry of the critical dynamics in zero field is also explored. Finally, we report some very recent 3-D Monte Carlo simulation results, which have been obtained by applying Multi-Spin-Coding techniques on a CDC vector supercomputer - Cyber 205 at John von Neumann Center.
Steady-state current transfer and scattering theory.
Ben-Moshe, Vered; Rai, Dhurba; Skourtis, Spiros S; Nitzan, Abraham
2010-08-01
The correspondence between the steady-state theory of current transfer and scattering theory in a system of coupled tight-binding models of one-dimensional wires is explored. For weak interwire coupling both calculations give nearly identical results, except at singular points associated with band edges. The effect of decoherence in each of these models is studied using a generalization of the Liouville-von Neuman equation suitable for steady-state situations. An example of a single impurity model is studied in detail, leading to a lattice model of scattering off target that affects both potential scattering and decoherence. For an impurity level lying inside the energy band, the transmission coefficient diminishes with increasing dephasing rate, while the opposite holds for impurity energy outside the band. The efficiency of current transfer in the coupled wire system decreases with increasing dephasing. PMID:20707524
Theory of minimum dissipation of energy for the steady state
The magnetic configuration of an inductively driven steady-state plasma bounded by a surface (or two adjacent surfaces) on which B·n = 0 is force-free: ∇xB = 2αB, where α is a constant, in time and in space. α is the ratio of the Poynting flux to the magnetic helicity flux at the boundary. It is also the ratio of the dissipative rates of the magnetic energy to the magnetic helicity in the plasma. The spatial extent of the configuration is noninfinitesimal. This global constraint is a result of the requirement that, for a steady-state plasma, the rate of change of the vector potential, ∂A/∂t, is constant in time and uniform in space
Steady-state propagation of interface corner crack
Veluri, Badrinath; Jensen, Henrik Myhre
2013-01-01
Steady-state propagation of interface cracks close to three-dimensional corners has been analyzed. Attention was focused on modeling the shape of the interface crack front and calculating the critical stress for steady-state propagation of the crack. The crack propagation was investigated by...... estimating the fracture mechanics parameters that includes the strain energy release rate, crack front profiles and the three-dimensional mode-mixity along the interface crack front. A numerical approach was then applied for coupling the far field solutions based on the Finite Element Method to the near...... field (crack tip) solutions based on the J-integral. The adopted two-dimensional numerical approach for the calculation of fracture mechanical properties was compared with three-dimensional models for quarter-circular and straight sided crack front shapes. A quantitative approach was formulated based on...
Steady states of the parametric rotator and pendulum
Bouzas, Antonio O
2011-01-01
We discuss several steady-state rotation and oscillation modes of the planar parametric rotator and pendulum with damping. We consider a general elliptic trajectory of the suspension point for both rotator and pendulum, for the latter at an arbitrary angle with gravity, with linear and circular trajectories as particular cases. We treat the damped, non-linear equation of motion of the parametric rotator and pendulum perturbatively for small parametric excitation and damping, although our perturbative approach can be extended to other regimes as well. Our treatment involves only ordinary second-order differential equations with constant coefficients, and provides numerically accurate perturbative solutions in terms of elementary functions. Some of the steady-state rotation and oscillation modes studied here have not been discussed in the previous literature. Other well-known ones, such as parametric resonance and the inverted pendulum, are extended to elliptic parametric excitation tilted with respect to gravi...
Master equation based steady-state cluster perturbation theory
Nuss, Martin; Dorn, Gerhard; Dorda, Antonius; von der Linden, Wolfgang; Arrigoni, Enrico
2015-09-01
A simple and efficient approximation scheme to study electronic transport characteristics of strongly correlated nanodevices, molecular junctions, or heterostructures out of equilibrium is provided by steady-state cluster perturbation theory. In this work, we improve the starting point of this perturbative, nonequilibrium Green's function based method. Specifically, we employ an improved unperturbed (so-called reference) state ρ̂S, constructed as the steady state of a quantum master equation within the Born-Markov approximation. This resulting hybrid method inherits beneficial aspects of both the quantum master equation as well as the nonequilibrium Green's function technique. We benchmark this scheme on two experimentally relevant systems in the single-electron transistor regime: an electron-electron interaction based quantum diode and a triple quantum dot ring junction, which both feature negative differential conductance. The results of this method improve significantly with respect to the plain quantum master equation treatment at modest additional computational cost.
Steady State Dynamic Operating Behavior of Universal Motor
Muhammad Khan Burdi
2015-01-01
Full Text Available A detailed investigation of the universal motor is developed and used for various dynamic steady state and transient operating conditions of loads. In the investigation, output torque, motor speed, input current, input/output power and efficiency are computed, compared and analyzed for different loads. While this paper discusses the steady-state behavior of the universal motor, another companion paper, ?Transient dynamic behavior of universal motor?, will discuss its transient behavior in detail. A non-linear generalized electric machine model of the motor is considered for the analysis. This study was essential to investigate effect of output load on input current, power, speed and efficiency of the motor during operations. Previously such investigation is not known
Steady-state Physics, Effective Temperature Dynamics in Holography
Kundu, Arnab
2013-01-01
Using the gauge-gravity duality, we argue that for a certain class of out-of-equilibrium steady-state systems in contact with a heat bath at a given temperature, the macroscopic physics can be captured by an effective thermodynamic description. The steady-state is obtained by applying a constant electric field that results in a stationary current flow. Within holography, we consider generic probe systems where an open string equivalence principle and an open string metric govern the effective thermodynamics. This description comes equipped with an effective temperature, which is larger than the bath temperature, and a corresponding effective entropy. For conformal or scale-invariant theories, certain scaling behaviours follow immediately. In general, in the large electric field limit, this effective temperature is also observed to obey certain generic relations with various physical parameters in the system.
Persistent Probability Currents in Non-equilibrium Steady States
Zia, Royce; Mellor, Andrew; Mobilia, Mauro; Fox-Kemper, Baylor; Weiss, Jeffrey
For many interesting phenomena in nature, from all life forms to the global climate, the fundamental hypothesis of equilibrium statistical mechanics does not apply. Instead, they are perhaps better characterized by non-equilibrium steady states, evolving with dynamical rules which violate detailed balance. In particular, such dynamics leads to the existence of non-trivial, persistent probability currents - a principal characteristic of non-equilibrium steady states. In turn, they give rise to the notion of 'probability angular momentum'. Observable manifestations of such abstract concepts will be illustrated in two distinct contexts: a heterogeneous nonlinear voter model and our ocean heat content. Supported in part by grants from the Bloom Agency (Leeds, UK) and the US National Science Foundation: OCE-1245944. AM acknowledges the support of EPSRC Industrial CASE Studentship, Grant No. EP/L50550X/1.
Steady state test on PWR steam generator thermohydraulics
Experimental activity on U-tube steam generator thermal hydraulics is under way at CISE and SIET in the framework of ENEA's LWR safety research programme. The test section includes 9 tubes. Hot side and cold side can be separated simulated, with primary and secondary fluid in full thermalhydraulic conditions. The experimental matrix includes: steady state tests (in both adiabatic and diabatic conditions); transients tests that simulate various accidents. Some steady state tests are reported. The secondary side average density, measured by the quick closing valve technique can be accurately calculated by the Zuber-Dix and Zuber-Rohuani correlations. Continuous pressure drops can be very well predicted by an adapted version of Thom correlation and CISE DIF-3 correlation: the development of an empirical correlation was, instead, necessary for assessment of the local pressure drops across spacer grids
Turnover of messenger RNA: Polysome statistics beyond the steady state
Valleriani, A.; Ignatova, Z.; Nagar, A.; Lipowsky, R.
2010-03-01
The interplay between turnover or degradation and ribosome loading of messenger RNA (mRNA) is studied theoretically using a stochastic model that is motivated by recent experimental results. Random mRNA degradation affects the statistics of polysomes, i.e., the statistics of the number of ribosomes per mRNA as extracted from cells. Since ribosome loading of newly created mRNA chains requires some time to reach steady state, a fraction of the extracted mRNA/ribosome complexes does not represent steady state conditions. As a consequence, the mean ribosome density obtained from the extracted complexes is found to be inversely proportional to the mRNA length. On the other hand, the ribosome density profile shows an exponential decrease along the mRNA for prokaryotes and becomes uniform in eukaryotic cells.
Amri, Amina; Pulko, Susan Helen; Wilkinson, Anthony James
2016-01-01
Breast thermography still has inherent limitations that prevent it from being fully accepted as a breast screening modality in medicine. The main challenges of breast thermography are to reduce false positive results and to increase the sensitivity of a thermogram. Further, it is still difficult to obtain information about tumour parameters such as metabolic heat, tumour depth and diameter from a thermogram. However, infrared technology and image processing have advanced significantly and recent clinical studies have shown increased sensitivity of thermography in cancer diagnosis. The aim of this paper is to study numerically the possibilities of extracting information about the tumour depth from steady state thermography and transient thermography after cold stress with no need to use any specific inversion technique. Both methods are based on the numerical solution of Pennes bioheat equation for a simple three-dimensional breast model. The effectiveness of two approaches used for depth detection from steady state thermography is assessed. The effect of breast density on the steady state thermal contrast has also been studied. The use of a cold stress test and the recording of transient contrasts during rewarming were found to be potentially suitable for tumour depth detection during the rewarming process. Sensitivity to parameters such as cold stress temperature and cooling time is investigated using the numerical model and simulation results reveal two prominent depth-related characteristic times which do not strongly depend on the temperature of the cold stress or on the cooling period. PMID:26522612
A Series RCL Circuit Theory for Analyzing Non-Steady-State Water Uptake of Maize Plants
Zhuang, Jie; Yu, Gui-Rui; Nakayama, Keiichi
2014-10-01
Understanding water uptake and transport through the soil-plant continuum is vital for ecosystem management and agricultural water use. Plant water uptake under natural conditions is a non-steady transient flow controlled by root distribution, plant configuration, soil hydraulics, and climatic conditions. Despite significant progress in model development, a mechanistic description of transient water uptake has not been developed or remains incomplete. Here, based on advanced electrical network theory (RLC circuit theory), we developed a non-steady state biophysical model to mechanistically analyze the fluctuations of uptake rates in response to water stress. We found that the non-steady-state model captures the nature of instantaneity and hysteresis of plant water uptake due to the considerations of water storage in plant xylem and coarse roots (capacitance effect), hydraulic architecture of leaf system (inductance effect), and soil-root contact (fuse effect). The model provides insights into the important role of plant configuration and hydraulic heterogeneity in helping plants survive an adverse environment. Our tests against field data suggest that the non-steady-state model has great potential for being used to interpret the smart water strategy of plants, which is intrinsically determined by stem size, leaf size/thickness and distribution, root system architecture, and the ratio of fine-to-coarse root lengths.
Non-equilibrium steady states for chains of four rotors
Cuneo, Noé; Eckmann, Jean-Pierre
2015-01-01
We study a chain of four interacting rotors (rotators) connected at both ends to stochastic heat baths at different temperatures. We show that for non-degenerate interaction potentials the system relaxes, at a stretched exponential rate, to a non-equilibrium steady state (NESS). Rotors with high energy tend to decouple from their neighbors due to fast oscillation of the forces. Because of this, the energy of the central two rotors, which interact with the heat baths only through the external ...
Steady-State Oscillations in Resonant Electrostatic Vibration Energy Harvesters
Blokhina, Elena; Galayko, Dimitri; Basset, Philippe; Feely, Orla
2013-01-01
In this paper, we present a formal analysis and description of the steady-state behavior of an electrostatic vibration energy harvester operating in constant-charge mode and using different types of electromechanical transducers. The method predicts parameter values required to start oscillations, allows a study of the dynamics of the transient process, and provides a rigorous description of the system, necessary for further investigation of the related nonlinear phenomena and for the optimis...
Simulation of Power Electronic Converters Using Quasi Steady State Approximation
Predrag Pejović
2012-01-01
A new method to compute voltage and current waveforms of power electronic converters is proposed in the paper. The method relies on simulation result of averaged circuit model, and superimposes the ripple of the inductor currents to the obtained average values, assuming that the linear ripple approximation applies. To determine the amplitude of the switching ripple, a quasi steady state approximation is proposed. After the inductor currents are obtained, currents of switching components are c...
Steady State Plasma Accelerators and their Applications in Thermonuclear Research
Steady state plasma accelerators make it possible in principle to obtain plasma fluxes of high energy and large flow rate. This is of interest in thermonuclear research for two reasons. Firstly, the accelerator can be used for injecting plasma into existing traps; secondly, it can be used to design new-types of thermonuclear reactors, which might be referred to as, ''Flow-type Reactors'' in which a positive yield is-realised during the time the material passes through the reactor system. The main types of accelerating mechanisms operating in this accelerator are described and a brief review is given of theoretical, numerical and experimental investigations carried out by the author and his colleagues. The numerical and theoretical analysis of the processes taking place in coaxial steady state accelerators revealed the possible existence of steady-state compressive flows during which the applied electromagnetic energy is not converted into kinetic plasma energy but is used for compression of the plasma. When the compressed flow is allowed to expand its thermal energy is converted into kinetic energy. Devices in which compressive flow is attained are referred to as magnetic plasma compressors. At the present stage the existence of compressive flows has been confirmed experimentally. To ensure a positive yield in-the region of compression a density of 1020 - 1020 cm-3 is essential. The possibility of obtaining a positive yield in linear ''Flow type Reactors'' is discussed. Such reactors consist of magnetic plasma guides of length ∼ 100 m, in which a flow of hot plasma is produced by a steady state plasma accelerator. (author)
Overview of recent experimental results from the DIII-D advanced tokamak program
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)