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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.

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.

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.

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.

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...

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.

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.

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.

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.

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.

A series of electron cyclotron heating (ECH) experiments was performed with a 28-GHz gyrotron on the Impurity Study Experiment (ISX-B) tokamak at Oak Ridge National Laboratory. Up to 70 kW of microwave power was injected into ISX-B from the high field side. Bulk heating was observed with a central temperature rise of approx. 370 eV from an original temperature of approx. 600 eV, as measured by Thomson scattering. With ECH and under low density conditions, large nonthermal signals were observed on electron cyclotron emission diagnostics at the first, second, and third harmonics. These signals sometimes became quite large after the end of the ECH pulse. The effects observed can be attributed to relatively small changes in the electron distribution function. The temporal behavior of the enhanced emission is tentatively attributed to the pitch angle scattering of superthermal electrons.

One experiment on electron cyclotron heating on the Impurity Study Experiment (ISX-b) tokamak has been completed and a second experiment is in progress. In the first experiment, with a gyrotron producing 100 kW at 35 GHz, a heating efficiency greater than 60% was achieved. Unpolarized radiation was launched from the high field side of the tokamak, heating electrons at cyclotron resonance in the center of the plasma. From laser scattering and thermal emission at the second gyroharmonic, the electron temperature, initially 850 eV, increased to 1250 eV at the end of a 15-ms pulse. The second experiment is now in progress, with 180 kW at 28 GHz and a capability of 100-ms pulses. A comparison between polarized and unpolarized radiation, modification of the plasma current profile, and a critical comparison with theory are planned. In this paper, experiments on preionization at the upper hybrid resonance are reported. A plasma with an electron temperature of 10 eV was produced, which is not high enough to reduce impurity radiation. The resistive part of the loop voltage was reduced by preionization. Plans for an experiment to measure electric current induced by electron cyclotron heating are presented, along with a review of the theoreticalmore » basis of the current drive.« less

A summary of experimental results pertaining to plasma energy and particle transport in the Impurity Study Experiment (ISX-B) tokamak is presented. Emphasis is placed on results with neutral beam heating, usually in the co- direction (aligned with the plasma current), and the relative roles of various energy loss mechanisms are discussed. The derived electron thermal diffusivity and the predictions of various models have been compared. The measured values are within a factor of 2 of the values expected to result from resistive pressure-driven modes. Evidence for the presence of these modes is discussed. Values of ion thermal diffusivity are compared with the predictions of neoclassical theory. Anomaly factors (the ratio of experimental value to theoretical value) between 1 and 5 are found. Comparisons between experimental results and theoretical predictions are also made for cases with increased toroidal field ripple, produced by using only 9 of the 18 toroidal field coils. Convection is shown to play a small role in energy transport, except at the plasma periphery. The behavior of both plasma and impurity particles is discussed and shown to be strongly dependent on the direction of the plasma current relative to the neutral beam (coinjection and counterinjection). Beta limits aremore » discussed. The maximum values obtained for both ..beta../sub I/ (approx. 2) and (approx. 2.5%) are not thought to be limited except by the restricted power available for heating.« less

A phased program of plasma heating at the electron cyclotron frequency is proposed for the Oak Ridge tokamaks ISX and ORMAK Upgrade. The past history of the program of electron cyclotron heating (ECH) at ORNL on mirrors and in the ELMO Bumpy Torus has been successful. Future technological developments in the production of high power high frequency microwave tubes look promising at this time. The physics of wave propagation and particle heating are fairly well understood and indicate the viability of this technique. Studies on breakdown and on runaway electron reduction will provide useful information for larger machines. Recent experiments in the USSR on small tokamaks have shown that ECH is a viable heating technique. Providing that the microwave tubes become available, the engineering considerations suggest that the technique is practical and workable, based on present day technology.

A high power ion Bernstein wave heating system has been introduced on PBX‐M for heating and for controlling the plasma pressure profile in an effort to achieve the stable high beta ‘‘second stability’’ regime. The pressure profile can be controlled through local bulk ion heating as well as density profile control. In bean‐shaped plasmas with plasma currents range from 180 kA to 250 kA, good ion heating up to the highest applied rf power, (≊700 kW,) has been observed. The observed broadening of the ion temperature profile is consistent with localized off‐axis bulk ion heating as predicted by IBW ray tracing calculations. Application of IBW also resulted in a greatly modified density profile. The ability for IBW to change the density profile appears to be particularly attractive for controlling the bootstrap current profile for advanced tokamaks. Many important IBWH‐related edge physics results were also obtained, including ponderomotive edge plasma modification and parametric instability onset conditions. ...

Recent experiments in the Advanced Toroidal Facility (ATF) torsatron [Plasma Physics and Controlled Nuclear Fusion Research 1990 (IAEA, Vienna, in press)] have emphasized the role of magnetic configuration control in transport studies. Long‐pulse plasma operation up to 20 sec has been achieved with electron cyclotron heating (ECH). With neutral beam injection (NBI) power of ≥1 MW, global energy confinement times of 30 msec have been obtained with line‐average densities up to 1.3×1020 m−3. The energy confinement and the operational space in ATF are roughly the same as those in tokamaks of similar size and field. The empirical scaling observed is similar to gyro‐reduced Bohm scaling with favorable dependences on density and field offsetting an unfavorable power dependence. The toroidal current measured during ECH is identified as the bootstrap current. The observed currents agree well with predictions of neoclassical theory in magnitude and in parametric dependence. Variations of the magnetic configuration ...