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We have observed reduction of the runaway electron synchrotron radiation, hard X-ray (HXR) intensity, and HXR energy after applying 110 GHz 2nd harmonic electron cyclotron resonant heating (ECRH) during runaway electron (RE) discharges at low density with startup runaway electrons. However, we did not see a significant reduction of X-rays from 170 GHz 2nd harmonic ECRH at a higher field. A recently installed IR TV camera was used to observe the forward cone of synchrotron radiation from high energy REs in KSTAR. We have observed changes to the synchrotron images and reduction of the HXR by application of resonant magnetic perturbations (RMP) from in-vessel control coils (IVCC) installed inside KSTAR in the n = 1 configuration.

A compact diagnostic system for monitoring Hard X-Ray (HXR) in the KSTAR tokamak has been developed in this work. Its development aims for an extension of the counting rate limit, and a good time r...

Wave enhanced runaway generation is expected to play an important role in the conversion of plasma current into runaway current during major disruptions. The fast electrons created by electron cyclotron heating (ECH) were used to study this issue in KSTAR. It is found that the fast electrons driven by ECH can enhance runaway production in the flat top phase with high loop voltage. The runaway current in disruptions was not enhanced by the ECH produced fast electron population due to the strong magnetic fluctuations which inhibited the generation of runaway electrons. It is found that a complete loss of existing REs during thermal quench has occurred in KSTAR limiter configuration discharges.

It is possible to detect the presence of small field errors in a tokamak with an electron beam. This was demonstrated earlier on T-15 and TEXTOR. This paper discusses the concept, past experience on these tokamaks, calculations for the Korea Superconducting Tokamak Advanced Research (KSTAR) device, an electron beam source, measurement devices for these measurements, and some results. It is shown that small toroidally averaged field errors can be detected by this method. A low voltage electron beam (e-beam) gun and fluorescent screen were mounted in a vertical port and inserted into the vacuum vessel at the end of the KSTAR 2nd campaign plasma experiments. A camera with a narrow field of view was mounted in midplane port in a tube tangent to the field lines at R ∼ 1.3 m and photographed the beam striking the screen. The poloidal field (PF) currents were held constant during the camera exposure period. Many shots with various PF coils energized were made and the deflections of the e-beam were measured. The measurements were made with a camera integration time of 300 ms because of the low light intensity. The results show that there are large field errors that diminish as the PF currents are raised. There appears to be no significant up–down asymmetry for static fields. Measurements with a 7 PF coil scenario with a calculated field null located at e-beam radial position show much larger fields than calculated. KSTAR was constructed with Incoloy 908 conduit using cable-in-conduit conductors (CICC) in 10 of the 14 PF coils and all 16 of the toroidal field (TF) coils. Incoloy 908 has a relative magnetic permeability, μ, of about 10. The field errors appear to be largely due to Incoloy 908.

There is strong evidence that the ICRH applied to the central cell (CC) of Hanbit does not result in complete ionization. Two techniques are being tried in an attempt to improve the degree of ionization. The first of them is ECRH preionization by application of microwave power from one 2-kW CPI klystron at 14 GHz as well as two 1.8-kW VA-806 klystrons at 7.67 GHz and 7.87 GHz. The second technique is production of a reflex discharge using a 76 mm diameter LaB{sub 6} cathode heated to incandescence and biased negatively in the cusp end cell of Hanbit. The ECRH preionization has shown beneficial effects on the plasma. There are no results from the reflex discharge.

Low density slide away discharges with anomalous Doppler resonance (ADR) effects have been observed in the KSTAR tokamak. When the line averaged electron density was lower than 0.6 × 10 19 m − 3 , the discharges went into the slide-away regime with relaxations in the electron cyclotron emission due to the ADR effects which transferred the runaway electron energy from parallel to perpendicular motion. The suppression of the ADR effects has been achieved by electron cyclotron resonance heating which enhanced the perpendicular energy of electrons and led to an isotropization of the electron distribution function.

New theoretical and experimental tools are applied to the analysis of ICRF antenna-edge plasma interactions in the TFTR tokamak. A new numerical method for computing the three dimensional (3-D) rf sheath voltage distribution is used, and the quantitative predictions of rf sheath theory are compared with measurements of the edge density profile obtained by microwave reflectometry and with titanium impurity concentration data. It is shown that the local density depletion at the antenna is consistent with density pump-out by strong E × B convection into the Faraday screen (FS). Modelling of the FS impurity influx shows that the calculated titanium impurity concentration based on this direct influx agrees with the measured concentration for π phasing. It is also shown that screening of impurity neutrals by ionization in the SOL is a large effect and increases with rf power. At high power over many shots, a fraction of the metal impurities migrates around the machine and is deposited on the limiters, providing a secondary source of titanium. The data show that the central titanium concentration is strongly dependent on antenna phasing. Possible explanations for this phasing dependence are discussed.

Continuing the experiments reported previously, 1 , 2 additional microwave power has been applied to the plug region of Hanbit in order to increase the stored energy and beta of the hot-electron plasma created there. Two new 1.5-kW VA-806 klystrons at 7.67 GHz and 7.87 GHz have been used in conjunction with the existing 2-kW CPI klystron at 14 GHz. The plasma is created in order to provide a high-beta ring to stabilize the Hanbit central cell plasma against ballooning instabilities. An array of Hall probes mounted on the outside of the Hanbit plug cavity was installed to measure the axial profile of the Bz fields. The total stored energy was measured by diamagnetic loops and the radial location of the plasma was determined by a Si-PIN diode detector measuring the energetic electron end loss. All three measurements were to be used to determine the radial and axial location of the plasma, the plasma volume, the stored energy, and hence the plasma beta. However, the Bz signal was too small to measure and the diamagnetic signal was smaller than previously found. The ring was found to be very wide and not adequate to stabilize the central cell plasma.

The three dimensional (3-D) nature of antennas for fusion applications in the ion cyclotron range of frequencies (ICRF) requires accurate modelling to design and analyse new antennas. In this article, analysis and design tools for radiofrequency (RF) antennas are successfully benchmarked with experiment, and the 3-D physics of the launched waves is explored. The systematic analysis combines measured density profiles from a reflectometer system, transmission line circuit modelling, detailed 3-D magnetostatics modelling and a new 3-D electromagnetic antenna model including plasma. This analysis gives very good agreement with measured loading data from the Tokamak Fusion Test Reactor (TFTR) Bay-M antenna, thus demonstrating the validity of the analysis for the design of new RF antennas. The 3-D modelling is contrasted with 2-D models, and significant deficiencies are found in the latter. The 2-D models are in error by as much as a factor of 2 in real and reactive loading, even after they are corrected for the most obvious 3-D effects. Three dimensional effects play the most significant role at low parallel wavenumbers, where the launched power spectrum can be quite different from the predictions of 2-D models. Three dimensional effects should not be ignored for many RF designs, especially those intended for fast wave current drive.

Density fluctuations in low-collisionality, low-beta ({beta} nearly equals 0.1%), currentless plasmas produced with electron cyclotron heating (ECH) in the Advanced Toroidal Facility (ATF) torsatron have been studied using a 2-mm microwave scattering diagnostic. Pulsed gas puffing is used to produce transient steepening of the density profile from its typically flat shape; this leads to growth in the density fluctuations when the temperature and density gradients both point in the same direction in the confinement region. The wave number spectra of the fluctuations that appear during this perturbation have a maximum at higher {kappa}{sub PRP}{rho}{sub s} (nearly 1) than is typically seen in tokamaks. The in-out asymmetry of the fluctuations along the major radius correlates with the distribution of confined trapped particles expected for the ATF magnetic field geometry. During the perturbation, the relative level of the density fluctuations in the confinement region (integrated over normalized minor radii {rho} from 0.5 to 0.85) increases from n/n >= 1% when the density profile flat to n/n >= 3% when the density profile is steepened. These observations are in qualitative agreement with theoretical expectations for helical dissipative trapped electron modes (DTEMs), which are drift-wave instabilities associated with particle trapping in the helical stellaratormore » field. 5 refs., 5 figs.« less

Application of Ion Bernstein Wave Heating (IBWH) into the Princeton Beta Experiment‐Modification (PBX‐M) [Phys. Fluids B 2, 1271 (1990)] tokamak stabilizes sawtooth oscillations and generates peaked density profiles. A transport barrier, spatially correlated with the IBWH power deposition profile, is observed in the core of IBWH‐assisted neutral beam injection (NBI) discharges. A precursor to the fully developed barrier is seen in the soft x‐ray data during edge localized mode (ELM) activity. Sustained IBWH operation is conducive to a regime where the barrier supports large ∇ne, ∇Te, ∇νφ, and ∇Ti, delimiting the confinement zone. This regime is reminiscent of the H(high) mode, but with a confinement zone moved inward. The core region has better than H‐mode confinement while the peripheral region is L(low)‐mode‐like. The peaked profile enhances NBI core deposition and increases nuclear reactivity. An increase in central Ti results from χi reduction (compared to the H mode) and better beam penetration. Boot...

Density fluctuations in low‐collisionality, low‐beta (β∼0.1%), currentless plasmas produced with electron cyclotron heating (ECH) in the Advanced Toroidal Facility (ATF) torsatron [Fusion Technol. 10, 179 (1986)] have been studied using a 2 mm microwave scattering diagnostic. Pulsed gas puffing is used to produce transient steepening of the density profile from its typically flat shape; this leads to growth in the density fluctuations when the temperature and density gradients both point in the same direction in the confinement region. The wave number spectra of the fluctuations that appear during this perturbation have a maximum at higher k⊥ρs (∼1) than is typically seen in tokamaks. The in–out asymmetry of the fluctuations along the major radius correlates with the distribution of confined trapped particles expected for the ATF magnetic field geometry. During the perturbation, the relative level of the density fluctuations in the confinement region (integrated over normalized minor radii ρ from 0.5 to 0...

A new type of reflectometer not previously used on tokamaks has been installed on TFTR to measure edge electron density profiles. This reflectometer gives unambiguous edge profiles despite large edge density fluctuations and the presence of strong auxiliary heating. The effects of the large density fluctuations in the edge gradient region are overcome by reducing the multiplicity of fringes and eliminating phase excursions produced by scattering from density fluctuations at the reflecting layer. This is accomplished by using differential-reflectometry: the difference phase between two probing signals reflecting from cut-off layers separated by a distance much less than the correlation length of the density fluctuations is used. This system probes the TFTR plasma using the extraordinary mode (X-mode) with two signals swept from 90-118 GHz while maintaining a fixed difference frequency of 125 MHz between these signals. It has been used to obtain density profiles in the range of 1*1011 to 3*1013 cm-3 in high-field (4.5-4.9 T) full size (R0=2.62 m, a=0.96 m) TFTR plasmas. The reflectometer launcher is located in an ion cyclotron range of frequencies (ICRF) antenna and views the plasma through a small penetration in the centre of the Faraday shield. Initial measurements demonstrated that this technique is an effective way to measure the electron density profile in the edge gradient region.

Positive limiter biasing on the currentless Advanced Toroidal Facility (ATF) torsatron produces a significant increase in the particle confinement with no improvement in the energy confinement. Experiments have been carried out in 1-T plasmas with {approximately}400 kill of electron cyclotron heating ECM. Two rail limiters located at the last closed flux surface (LCFS), one at the top and one at the bottom of the device, are biased at positive and negative potentials with respect to the vessel. When the limiters are positively biased at up to 300 V, the density increases sharply to the ECH cutoff value. At the same time, the H{sub {alpha}} radiation drops, indicating that the particle confinement improves. When the density is kept constant, the H{sub {alpha}} radiation is further reduced and there is almost no change in the plasma stored energy. Under these conditions, the density profile becomes peaked and the electric field becomes outward-pointing outside the LCFS and more negative inside the LCFS. In contrast, negative biasing yields some reduction of the density and stored energy at constant gas feed, and the plasma potential profile remains the same. Biasing has almost no effect on the intrinsic impurity levels in the plasma.

Fast wave heating experiments in the ion cyclotron range of frequencies (ICRF) were performed on target plasmas produced by 350 kW of electron cyclotron heating at 53 GHz and also by neutral beam injection in the Advanced Toroidal Facility (ATF). Various heating regimes were investigated in the frequency range between 9.2 MHz and 28.8 MHz with magnetic fields of 0.95 T and 1.9 T on axis. The nominal pulse lengths of up to 200 kW RF power were in the range between 100 and 400 ms. Data from spectroscopy, loading measurements, and edge RF and Langmuir probes were used to characterize the RF induced effects on the ATF plasma. In the hydrogen minority regime at low plasma density, large suprathermal ion tails were observed with a neutral particle analyser. At high density (ne ≥ 5.0 × 1013 cm-3) substantial increases in antenna loading were observed, but ICRF power was insufficient to produce definitive heating results. A two-dimensional RF heating code, ORION, and a Fokker-Planck code, RFTRANS, were used to simulate these experiments. A simulation of future high power, higher density experiments in ATF indicates improved bulk heating results due to the improved loading and more efficient thermalization of the minority tail

Thermal plasma collapses have been observed in the Advanced Toroidal Facility (ATF) and in some cases these collapses limit the plasma performance. A 15-channel bolometer array has been used to measure the radiated power profile. It has been found that thermal collapses, defined by a drop in the plasma stored energy, occur when the local radiated power exceeds the local power input to the electrons. The local radiated power at the collapse first increases at the edge and then begins to propagate inward, leading to a peaking of the radiated power profile. This finding indicates that additional central heating would be effective in preventing the thermal collapse

Recent experiments in the Advanced Toroidal Facility (ATF) [Fusion Technol. 10, 179 (1986)] have been directed toward investigations of the basic physics mechanisms that control confinement in this device. Measurements of the density fluctuations throughout the plasma volume have provided indications for the existence of theoretically predicted dissipative trapped electron and resistive interchange instabilities. These identifications are supported by results of dynamic configuration scans of the magnetic fields during which the magnetic well volume, shear, and fraction of confined trapped particles are changed continuously. The influence of magnetic islands on the global confinement has been studied by deliberately applying error fields which strongly perturb the nested flux‐surface geometry, and the effects of electric fields have been investigated by means of biased limiter experiments.

Runaway electron formation and confinement occur readily in pulsed torsatrons and heliotrons because of the high loop voltages during initiation and termination of the helical and vertical fields (‘‘field ramping’’) and the inherently good containment of the electrons on the flux surfaces in the vacuum fields. This has been confirmed for the Advanced Toroidal Facility (AFT) [Fusion Technol. 10, 179 (1986)] and other stellarators by orbit calculations. Since runaway electrons can cause an unacceptable level of hard x rays near the machine, a runaway electron suppression system was incorporated in ATF. The main component of the system is a movable paddle, which is normally left in the center of the plasma chamber during the field ramps. This device, in conjunction with programmed vertical field ramping, which reduces the volume of good flux surfaces, has proved to be very effective in reducing the runaway electron population. Measurements of hard x rays from ATF have shown that the runaway electrons are pro...

The toroidal current observed during electron cyclotron heating in the Advanced Toroidal Facility torsatron is identified as bootstrap current. The observed currents, ranging between +4 and {minus}2 kA, agree well with predictions of neoclassical theory in magnitude and parametric dependence, as determined by systematic scans of quadrupole (shaping) and dipole (magnetic axis shift) moments of the poloidal magnetic field. It has been shown that the bootstrap current in a stellarator can be externally controlled, and zero-current operation can be achieved.

The characteristics of plasmas in the Advanced Toroidal Facility (ATF) have proven to be strongly dependent on the type of wall conditioning employed. A succession of techniques, beginning with glow discharge cleaning and baking, and evolving to gettering with chromium and titanium, have led to progressive improvement of the plasma parameters. Gettering with titanium has reduced the low-Z impurity content by a factor of 3, lowered the radiated power by a factor of 2.5–3.5, and improved the control over the electron density. The maximum values achieved for stored energy, line averaged density and confinement times are 28 kJ, 1.2 × 1014cm−3 and 25 ms, respectively. These parameters are comparable to the best results achieved in the ISX-B tokamak which had the same average minor radius and one half the major radius of ATF. Quasi-steady operation for 200 ms of neutral beam injection (NBI) has been obtained in high density, titanium gettered plasmas without the collapses that were typical earlier periods of operation. Neon injection experiments have helped to delineate the limits on the global levels of radiation that can be maintained and have supported the conclusion that mechanisms other than radiative losses are important for initiating the collapses still observed in low density NBI plasmas.

Plasmas in the Advanced Toroidal Facility (ATF), an l = 2 torsatron with 12 field periods, are produced by 200–400 kW of electron cyclotron heating (ECH) and up to 1.5 MW of neutral-beam injection (NBI). The characteristics of the plasmas are sensitive to the type of wall conditioning employed. A progression of techniques, beginning with glow discharge cleaning and baking and evolving to gettering with chromium and titanium, has significantly reduced the low-Z impurity content, lowered the fraction of radiated power, and permitted improved control over the electron density. As a result, plasma-parameters and machine performance have been enhanced significantly. The maximum values achieved for stored energy, line-averaged density, and confinement times are 28 kJ, 1.2 × 1020m−3, and 25 ms. These parameters are comparable to those obtained in the ISX-B tokamak. In addition, with titanium gettering, quasisteady operation for 200 ms of neutral beam injection has been obtained without the collapses that characterized earlier periods of operation.

A workshop on 1–3 August 1989 reviewed the techniques, uncertainties, and experiences of neutron calibration on PLT, TFTR, JET, Tore Supra, JT‐60, JIPPT‐IIU, Alcator C‐Mod, ATF, FT, ASDEX, Textor, and DIII‐D. In the summary session, the workshop participants discussed possible consensus neutron calibration techniques appropriate to D‐D plasmas in tokamaks. The application of such techniques would facilitate a more accurate comparison of neutron yields from different devices, and also allow new calibration techniques to relate their precision to a reference value. General agreement was reached on the suitability of two techniques: (1) a 252Cf source calibration of epithermal neutron detectors, and (2) threshold neutron activation of Ni foils placed vertically above or below the plasma. This paper will present details on detector positioning, neutron transport calculations, and interlab normalization needed to accomplish the standardized calibration using a Cf neutron source.

Experiments involving plasma improvement, confinement scaling, bootstrap currents, and edge fluctuations have been carried out in the Advanced Toroidal Facility (ATF) torsatron [Fusion Technol. 10, 179 (1986)]. Average densities ne≤9×1019 m−3 have been obtained, with global energy confinement times τ*E≤20 msec. Confinement times generally follow the stellarator/torsatron empirical scaling law, τSL =0.17×P−0.58n0.69eB0.84a2R0.75 (with τSL in seconds, power P in megawatts, density ne in 1020 m−3, and plasma radius a and major radius R in meters). Gas injection during neutral beam injection (NBI) causes increases in ne, so that τ*E does not decrease during NBI. Edge plasma fluctuations are found to exhibit a mode change near the peak of the energy confinement time. Plasma currents observed during electron cyclotron heating have been identified as bootstrap currents.

Edge electron density profile measurements, including the scrape‐off layer, have been made during ion cyclotron range of frequency (ICRF) heating with the two‐frequency differential‐phase reflectometer installed on an ICRF antenna on the Tokamak fusion test reactor (TFTR). This system probes the plasma using the extraordinary mode with two signals swept from 90 to 118 GHz, while maintaining a fixed‐difference frequency of 125 MHz. The extraordinary mode is used to obtain density profiles in the range of 1×1011–3×1013 cm−3 in high‐field (4.5–4.9 T) full‐size (R0=2.62 m, a=0.96 m) TFTR plasmas. The reflectometer launcher is located in an ICRF antenna and views the plasma through a small penetration in the center of the Faraday shield. A 26‐m‐long overmoded waveguide run connects the launcher to the reflectometer microwave electronics. Profile measurements made with this reflectometer system will be presented along with a discussion of the characteristics of this differential phase reflectometer and data ana...

Summary form only given, as follows. Initial attempts have been made at Ohmic (Joule) heating, stabilization, and preionization in Hanbit using an electron-emitting hot cathode in conjunction with a reflex discharge. The hot cathode was constructed using a set of LaB/sub 6/ discs heated internally with current through carbon rods. The cathode was mounted in the cusp section of Hanbit on magnetic field lines connected to the central cell (CC). The reflex discharge was initiated between the cathode and a CC biased limiter. This provided Joule heating of the electrons in the plasma Normally, ion cyclotron resonant heating (ICRH) produces the plasma. In the CC with a slot antenna supplied with power from a 500-kW transmitter at 3.5 MHz. In addition, the reflex discharge could be initiated prior to the ICRH discharge in order to provide a target plasma for the ICRH power. The ICRF power was thus not needed to breakdown the plasma and could be used solely for ion heating. Further experiments have been performed with the heated cathode but, without a reflex discharge to attempt line-tying stabilization of the CC plasma with all electron emitting end wall. The preliminary results for these three experiments will be presented. In addition, the electron-emitting cathode could be used to remove gas on the interior surfaces of the CC by electron impact desorption (EID) when the magnetic fields are arranged to cause the electrons to strike the CC walls arid internal structures. Preliminary results from this experiment will also be presented.

Summary form only given, as follows. Additional microwave power has been installed at Hanbit in an attempt to increase the beta of the hot-electron ring plasma created in the plug section. Two VA-806 series klystrons with a frequency of /spl sim/8 kW have been installed to operate in conjunction with an existing 2-kW 14-GHz klystron already, used in previous experiments. A hot-electron ring is created in the plug in order to provide a high-beta ring to stabilize the Hanbit central cell plasma against ballooning instabilities. Previous experiments had shown that a hot-electron ring plasma had been created with the 14-GHz klystron. The hot-electron ring was in a favorable location to stabilize the CC plasmas but the plasma beta was too low and could not provide stability. The present experiments were designed to increase the beta with additional microwave power. In addition, diagnostics were installed to determine the length of the plasma New diagnostic techniques and measurements will be presented.

In stellarators, external coil currents rather than internal currents are used to produce the confinement fields. This allows for greater flexibility in the design and operation of a stellarator. This flexibility can be exploited in a series of experiments on one device, such as the ATF. MHD stability boundaries will be tested by varying the vertical field. Access to the second stability regime has been observed in initial experiments on ATF. In this regime, the relationship between stabilization of pressure-driven modes and transport improvement can be studied. The density profile peaking factor can be varied using pellet injection in order to understand the role of eta /sub j/ modes. As an example, the ATF-TEXT collaboration on edge fluctuations will compare the levels of resistive-gradient-driven fluctuations observed in the two devices. In addition, global confinement and local transport comparisons have been made to tokamaks and to other stellarators of roughly similar size and capabilities but with widely varying configuration properties. >

Since the first H-mode discharges in 2010, the duration of the H-mode state has been extended and a significantly wider operational window of plasma parameters has been attained. Using a second neutral beam (NB) source and improved tuning of equilibrium configuration with real-time plasma control, a stored energy of Wtot???450?kJ has been achieved with a corresponding energy confinement time of ?E???163?ms. Recent discharges, produced in the fall of 2012, have reached plasma ?N up to 2.9 and surpassed the n?=?1 ideal no-wall stability limit computed for H-mode pressure profiles, which is one of the key threshold parameters defining advanced tokamak operation. Typical H-mode discharges were operated with a plasma current of 600?kA at a toroidal magnetic field BT?=?2?T. L?H transitions were obtained with 0.8?3.0?MW of NB injection power in both single- and double-null configurations, with H-mode durations up to ?15?s at 600?kA of plasma current. The measured power threshold as a function of line-averaged density showed a roll-over with a minimum value of ?0.8?MW at . Several edge-localized mode (ELM) control techniques during H-mode were examined with successful results including resonant magnetic perturbation, supersonic molecular beam injection (SMBI), vertical jogging and electron cyclotron current drive injection into the pedestal region. We observed various ELM responses, i.e. suppression or mitigation, depending on the relative phase of in-vessel control coil currents. In particular, with the 90? phase of the n?=?1 RMP as the most resonant configuration, a complete suppression of type-I ELMs was demonstrated. In addition, fast vertical jogging of the plasma column was also observed to be effective in ELM pace-making. SMBI-mitigated ELMs, a state of mitigated ELMs, were sustained for a few tens of ELM periods. A simple cellular automata (?sand-pile?) model predicted that shallow deposition near the pedestal foot induced small-sized high-frequency ELMs, leading to the mitigation of large ELMs. In addition to the ELM control experiments, various physics topics were explored focusing on ITER-relevant physics issues such as the alteration of toroidal rotation caused by both electron cyclotron resonance heating (ECRH) and externally applied 3D fields, and the observed rotation drop by ECRH in NB-heated plasmas was investigated in terms of either a reversal of the turbulence-driven residual stress due to the transition of ion temperature gradient to trapped electron mode turbulence or neoclassical toroidal viscosity (NTV) torque by the internal kink mode. The suppression of runaway electrons using massive gas injection of deuterium showed that runaway electrons were avoided only below 3?T in KSTAR. Operation in 2013 is expected to routinely exceed the n?=?1 ideal MHD no-wall stability boundary in the long-pulse H-mode (?10?s) by applying real-time shaping control, enabling n?=?1 resistive wall mode active control studies. In addition, intensive works for ELM mitigation, ELM dynamics, toroidal rotation changes by both ECRH and NTV variations, have begun in the present campaign, and will be investigated in more detail with profile measurements of different physical quantities by techniques such as electron cyclotron emission imaging, charge exchange spectroscopy, Thomson scattering and beam emission spectroscopy diagnostics.

The high fraction of runaway current conversion following disruptions has an important effect on the first wall for next-generation tokamaks. Because of the potentially severe consequences of a large full current runaway beam on the first wall in an unmitigated disruption, runaway suppression is given a high priority. The behavior of runaway currents both in spontaneous disruptions and in D2 massive gas injection (MGI) shutdown experiments is investigated in the KSTAR tokamak. The experiments in KSTAR show that the toroidal magnetic field threshold, BT >2 T, for runaway generation is not absolute. A high fraction of runaway current conversion following spontaneous disruptions is observed at a much lower toroidal magnetic field of BT = 1.3 T. A dedicated fast valve for high-pressure gas injection with 39.7 bar is developed for the study of disruptions. A study of runaway current parameters shows that the conversion efficiency of pre-disruptive plasma currents into runaway current can reach over 80% both in spontaneous disruptions and in D2 MGI shutdown experiments in KSTAR.

Ion cyclotron range of frequency (ICRF) heating is currently in use on CHS and W7‐AS and is a major element of the heating planned for steady state helical devices. In helical devices, the lack of a toroidal current eliminates both disruptions and the need for ICRF current drive, simplifying the design of antenna structures as compared to tokamak applications. However the survivability of plasma facing components and steady state cooling issues are directly applicable to tokamak devices. Results from LHD steady state experiments should be available on a time scale to strongly influence the next generation of steady state tokamak experiments. The helical plasma geometry provides challenges not faced with tokamak ICRF heating, including the potential for enhanced fast ion losses, impurity accumulation, limited access for antenna structures, and open magnetic field lines in the plasma edge. The present results and near term plans provide the basis for steady state ICRF heating of larger helical devices. An a...

The design, analysis, and performance evaluation of rf power systems ultimately requires accurate modeling of a chain of subsystems starting with the rf transmitter and ending with the power absorption in the plasma. A collection of computer codes is used at ORNL to calculate the plasma loading and wave spectrum for a three‐dimensional rf antenna, the transmission/reflection properties of the Faraday shield and its effect on the electrical characteristics and phase velocity of the antenna, the internal coupling among antenna array components and the incorporation of the antenna array into a transmission line model of the phase control, tuning, matching, and power distribution system. Some codes and techniques are more suited for the rapid evaluation of system design progressions, while others are more applicable to the detailed analysis of final designs or existing hardware. The interaction of codes and the accuracy of calculations will be illustrated by the process of determining the plasma loading as a function of phasing and density profiles for the TFTR ICRH antennas and comparing the results to measurements. An example of modeling a complex antenna geometry will be the comparison of calculations with the measured electrical response of a four‐strap mockup of the JET A2 antenna array which was loaned to ORNL by the JET ICRH team.

Since the successful first plasma generation in the middle of 2008, three experimental campaigns were successfully made for the KSTAR device, accompanied with a necessary upgrade in the power supply, heating, wall-conditioning and diagnostic systems. KSTAR was operated with the toroidal magnetic field up to 3.6 T and the circular and shaped plasmas with current up to 700 kA and pulse length of 7 s, have been achieved with limited capacity of PF magnet power supplies. The mission of the KSTAR experimental program is to achieve steady-state operations with high performance plasmas relevant to ITER and future reactors. The first phase (2008–2012) of operation of KSTAR is dedicated to the development of operational capabilities for a super-conducting device with relatively short pulse. Development of start-up scenario for a super-conducting tokamak and the understanding of magnetic field errors on start-up are one of the important issues to be resolved. Some specific operation techniques for a super-conducting device are also developed and tested. The second harmonic pre-ionization with 84 and 110 GHz gyrotrons is an example. Various parameters have been scanned to optimize the pre-ionization. Another example is the ICRF wall conditioning (ICWC), which was routinely applied during the shot to shot interval. The plasma operation window has been extended in terms of plasma beta and stability boundary. The achievement of high confinement mode was made in the last campaign with the first neutral beam injector and good wall conditioning. Plasma control has been applied in shape and position control and now a preliminary kinetic control scheme is being applied including plasma current and density. Advanced control schemes will be developed and tested in future operations including active profiles, heating and current drives and control coil-driven magnetic perturbation.

This letter reports on the successful demonstration of the second harmonic electron cyclotron heating (ECH)-assisted startup in the first plasma experiments recently completed in the fully superconducting Korea Superconducting Tokamak Advanced Research (KSTAR) device whose major and minor radii are 1.8 m and 0.5 m, respectively. For the second harmonic ECH-assisted startup, an 84 GHz EC wave at 0.35 MW was launched before the onset of the toroidal electric field of the Ohmic system. And it was observed that this was sufficient to achieve breakdown in the ECH pre-ionization phase, allow burn-through and sustain the plasma during the current ramp with a low loop voltage of 2.0 V and a corresponding toroidal electric field of 0.24 V m−1at the innermost vacuum vessel wall (R = 1.3 m). This is a lower value than 0.3 Vm−1 which is the maximum electric field in ITER. Due to the limited volt-seconds and the loop voltage of the Ohmic power system, the extended pulse duration of the ECH power up to 180 ms allowed the plasma current to rise up to more than 100 kA with a ramp-up rate of 0.8 MA s−1.

An overview of recent experimental results from the Advanced Toroidal Facility (ATF) is presented. Beam-heated plasmas with {bar n}{sub e} of 10{sup 20} m{sup {minus}3} and {tau}{sub E}{asterisk} of {approx}20 ms have been achieved. Thermal collapse of the plasmas is mitigated by wall conditioning and particle fueling. Confinement time scales positively with density and magnetic field, offsetting deterioration with power. Results fit the LHD scaling and the drift wave turbulence scaling. Bootstrap currents observed during ECH agree with neoclassical theory in magnitude and parameter dependences. Fast reciprocating Langmuir probe measurements show that edge fluctuations in ATF have many similarities to those in the TEXT tokamak. The location of {tilde B} instabilities has shifted outward in radius, consistent with the broader pressure profiles. 14 refs., 6 figs.

A workshop on 1–3 August 1989 reviewed the techniques, uncertainties, and experiences of neutron calibration on PLT, TFTR, JET, Tore Supra, JT‐60, JIPPT‐IIU, Alcator C‐Mod, ATF, FT, ASDEX, Textor, and DIII‐D. In the summary session, the workshop participants discussed possible consensus neutron calibration techniques appropriate to D‐D plasmas in tokamaks. The application of such techniques would facilitate a more accurate comparison of neutron yields from different devices, and also allow new calibration techniques to relate their precision to a reference value. General agreement was reached on the suitability of two techniques: (1) a 252Cf source calibration of epithermal neutron detectors, and (2) threshold neutron activation of Ni foils placed vertically above or below the plasma. This paper will present details on detector positioning, neutron transport calculations, and interlab normalization needed to accomplish the standardized calibration using a Cf neutron source.

An experiment to test beryllium as a limiter material has been performed in the ISX-B tokamak. The effect of the plasma on the limiter and the effect of the limiter on the plasma were studied in detail. Heat and particle fluxes to the limiter were measured, and limiter damage by melting was documented as a function of power flux. Strong melting and evaporation of the limiter caused beryllium gettering of the vacuum vessel. Postmortem analysis of the limiter was performed to document the amount of retained hydrogen and the erosion and impurity deposition on the limiter. The effect of the limiter on the plasma performance was studied in terms of parameter space, impurity content, and confinement for the ungettered and gettered cases. Operational experience with beryllium in a fusion experiment is discussed.

One of 12 beryllium tiles on a top rail limiter in the Impurity Study Experiment (ISX-B) tokamak was intentionally made radioactive. The migration of the radioactivity due to melting, ablation, and general excoriation of the radioactive tile was studied. Several hundred miligrams of material from the radioactive tile were spread to the other tiles over the course of the experiment, which consisted of more than 3000 tokamak discharges. A small amount of activity was found on the bottom of the tokamak vacuum vessel. The distribution of activity on the other limiter tiles showed a marked radially inward directivity, although all of the tiles received some activity. The possibility that some of the activity was the result of photo- and electro-excitation by runaway electrons cannot be ruled out but probably cannot account for the bulk of the effect.

RF power is widely used as an auxiliary heating method in fusion devices. This paper reviews the relevant theoretical considerations for the ion cyclotron, lower hybrid and electron cyclotron ranges of frequency, and presents the history, the state of the art, and the plans and prospects for antennas and transmission lines for RF heating. Reactor-relevant concerns are discussed, and the information needed to develop realistic antenna designs for a reactor environment is assessed.

A system has been developed for neutron flux measurements at the ORMAK tohamak device. This system allows measurement of the time-dependence of the neutron flux during a single shot of the device as well as the energy spectrum of the neutrons. Pulse shape discrimination techniques are used with a liquid scintillation counter (NE 213) to eliminate events in the detector due to gamma rays and X-rays generated in the plasma. Results are presented for the mode in which neutral deuterium is injected into a deuterium plasma.

Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge exchange neutral particle analyser. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and 3He ions, respectively. When the plasma was compressed, the d(d,n)3He fusion reaction rate increased by a factor of five, and the 3He(d,p)α rate by a factor of twenty. These data were simulated with a bounce averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling.

Measurements of temperature rise of the wall and of a movable limiter have been made in the Oak Ridge Tokamak (ORMAK). For a normal discharge without injection, approximately 70% of the energy in the discharge is deposited on the wall. Very little energy is deposited on a point limiter until it is moved several centimetres inside the circular limiter, indicating that the body of the plasma is not in contact with the limiter and that diffusion and heat conduction do not transfer a large fraction of the plasma energy to the limiter.

Initial plasmas have been obtained in the Advanced Toroidal Facility (ATF) at Oak Ridge. During the first operating period all major systems worked well. Currentless plasmas have been formed with electron-cyclotron heating (ECH) and neutral beam injection (NBI). Plasma breakdown is achieved with 200 kW of microwave power using the second harmonic of the 53 GHz gyrotron at 1 T. Because of the insufficient initial cleanup state of the vessel, the density evolution during the discharge is dominated by impurity influxes due to plasma-wall interactions. Wall conditioning has so far been performed with glow discharge and electron-cyclotron resonance discharge cleaning, with the walls at room temperature. Plasma edge studies are in preparation. They are focused on experiments leading to a divertor concept for ATF. During this brief operating period, plasma conditions were dominated by wall interactions. It is expected that during the next period additional wall conditioning techniques will lead to improved plasma performance.

Pre-ionization experiments have been performed on a tokamak by injecting about 80 kW of microwave power at 35 GHz for up to 15 ms. Microwave absorption occurs at the electron cyclotron and upper hybrid resonance frequencies as predicted by theory. Pre-ionization causes substantial (40%) reductions in loop voltage during the initial phase of the tokamak shot. Flux (volt-second) savings with pre-ionization are about 30% in the first 2 ms or about 2% of the total flux expenditure in a tokamak shot. The plasma current begins 200 μs earlier and rises 1.4 times more rapidly in the pre-ionized case. Electron densities of 5 × 1012 cm−3 can be sustained throughout the microwave pulse with only a toroidal magnetic field during microwave injection. The bulk electron temperature in the pre-ionized plasma is about 10 eV although there are indications of higher electron temperatures (50 eV) in the upper hybrid resonance layer. Although questions exist concerning the quiescent behaviour of the pre-ionized plasma, the observed parameters are shown to be consistent with a theory which employs classical models of energy and particle balance. During the early stages of Ohmic heating, the pre-ionization is effective in decreasing the peak of the radiated power.

In linear wave-particle interaction models, the coupling between cold electrons and microwaves with frequency equal to twice the electron gyrofrequency is so weak that the ionization of a significant number of neutral hydrogen atoms would seem impossible in practical applications. However, the non-linear interaction of a cold electron with the wave is very large if the electron becomes trapped near resonance in a shallow, static magnetic well. A model has been developed to describe the breakdown of a neutral gas when these non-linear interactions are considered, and it is in reasonable agreement with the limited amount of available experimental data. For gas pressures that are too large, electron-neutral collisions inhibit the non-linear interaction and prevent breakdown. For gas pressures that are too low, the growth rate of the free electron population is limited because electrons capable of causing ionization are lost before suffering a collision with a neutral. Quantitative growth rate predictions are presented for stellarators, and formulae for rough estimates are given.

The ISX-A was a tokamak designed for studying plasma-wall interactions and plasma impurities. It fulfilled this role quite well, producing reliable and reproducible plasmas which had currents up to 175 kA and energy containment times up to 30 ms. With discharge precleaning, Zeff was as low as 1.6; with titanium evaporation, Zeff approached 1.0. Values of Zeff≳ 2.0 were found to be proportional to residual impurity gases in the vacuum system immediately following a discharge. However, there was no clear dependence of Zeff on base pressure. Stainless steel limiters were used in most of the ISX-A experiments. Upon introducing carbon limiters into the vacuum system, Zeff increased to 5.6. After twelve days of clean-up with tokamak discharges, during which time Zeff steadily decreased, the carbon limiters tended to give slightly higher values of Zeff than stainless steel limiters. Injection of

Results are reported of electron-cyclotron-heating experiments in which 80 kW of microwave power from a 35-GHz gyrotron is injected into a tokamak with large single-pass absorption. For 10-ms microwave pulses, incident from the high-field side of the torus, the central electron temperature increases from 850 to 1250 eV, in agreement with empirical transport-code calculations. For the first time it is demonstrated that electron temperature in a tokamak scales linearly with electron-cyclotron-heating power.

A model is proposed for the movement and erosion of limiter material during the Beryllium Limiter Experiment performed on the ISX-B Tokamak. This model is consistent with observed experimental results and plasma operational characteristics. Conclusions drawn from the model can provide an understanding of erosion mechanisms, thereby contributing to the development of future design criteria.

The sequence of experiments resulting in the development of steady-state dense hot-electron plasmas is briefly described. These plasmas are produced in magnetic-mirror machines by radiation at the electron-cyclotron frequency. The electron-cyclotron plasma with the greatest stored energy to date has a volume of ~ 50 liters, an electron "temperature" of 120 keV, a density of 4 — 7 × 1011/cm3, an d an average beta of ~ 0.4. This plasma is created in the EPA Facility by a 50-kW c.w. 10.6-gHz microwave power source. The construction and operation of the machine are briefly discussed. The analysis and interpretation of the bremsstrahlung and diamagnetic measurements employed to determine these parameters are given. Diamagnetic and particle decay measurements on a smaller machine, the Physics Test Facility, are also described.

It is observed that the expression β = Aβqψ / κ = 18–22% defines one of the operational boundaries for ISX-B and corresponds to empirical beta limit scalings obtained in other tokamaks. Although this expression closely agrees with simple expressions for ideal-MHD instability thresholds, it is not obvious that it constitutes a real tokamak beta limit. Confinement degradation is observed when β is well below this operating boundary, and there is neither experimental evidence nor a satisfactory theoretical model connecting this phenomenon with ideal-MHD activity.

One of 12 beryllium tiles on a top rail limiter in the Impurity Study Experiment (ISX-B) tokamak was intentionally made radioactive. The migration of the radioactivity due to melting, ablation, and general excoriation of the radioactive tile was studied. Several hundred milligrams of material from the radioactive tile were spread to the other tiles over the course of the experiment, which consisted of more than 3000 tokamak shots. A small amount of activity was found on the bottom of the tokamak vacuum vessel. The distribution of activity on the other limiter tiles showed a marked radially inward directivity, although all of the tiles received some activity. The possibility that some of the activity was the result of photo- and electro-excitation by runaway electrons cannot be ruled out but probably cannot account for the bulk of the effect.