Beam-heated Discharges Research Articles

Stability of beta-induced Alfvén eigenmodes (BAE) in DIII-D

Although the stability of ellipticity, toroidal and reversed-shear Alfvén eigenmodes (EAE, TAE, RSAE) are relatively well understood, less is known about the stability of lower-frequency modes such as the beta-induced Alfvén eigenmode (BAE) but, because they are often unstable in present devices and are implicated in fast-ion transport, understanding their stability is vital. BAE stability is studied in primarily weak or reversed shear DIII-D plasmas with sub-Alfvénic deuterium beams. Modes are classified based on electron cyclotron emission, beam emission spectroscopy, magnetics, and interferometer data. The study is limited to the initial two seconds of the discharge, where the evolving q profile provides an effective scan of the dependence of stability upon q. In a dedicated experiment, BAEs are unstable at times in the discharge when the minimum of the safety factor q min is close to a rational number. The observed mode frequencies are usually close to analytic estimates of the BAE accumulation point and the eigenfunction peaks in the vicinity of q min. Unstable BAEs usually occur in bursts that chirp rapidly in frequency. To isolate the importance of thermal and beam gradients in driving the modes, the beam and electron cyclotron heating power is altered for 50–100 ms durations in reproducible discharges. As expected from the resonance condition, BAEs depend sensitively on the beam power and injection geometry. Modes only persist for ∼25 ms because the anisotropic beam population only interacts strongly with the modes over a relatively narrow range of q. A database of over 1000 beam-heated discharges shows that BAEs are more likely to be unstable when the poloidal beta exceeds 0.5.

Simulation Analysis of Charge-Exchange Losses During Neutral Beam Injection on EAST

Neutral beam injection (NBI) is one of the most effective heating methods on modern tokamak. As one of the two main losses during the NBI discharge, the charge-exchange (CX) losses are hard to measure due to their scattering. In this paper, through the simulation analysis with NUBEAM and ONETWO, CX losses have been investigated to obtain a rough evaluation by comparing to the shinethrough losses or the beam power. Firstly, according to the simulation results, the ratio of CX losses to the shinethrough losses varies from 0.28 to 1.87 in the scanning of plasma and NBI parameters. Thus, CX losses should not be ignored in the evaluations of the Low–High power threshold and beam torque in beam-heated discharges. Secondly, when Te(0) > 1 keV, CX losses have little relationship with the increasing temperature and can be roughly estimated as 17 % of the beam power or 85 % of the shinethrough losses. Finally, in the high collisional plasma, CX losses and shinethrough losses increase with the injected 80 keV neutral beam power. However, the ratio of them is almost similar. Thus, CX losses can be estimated as about 30 % of the shinethrough losses. On the other hand, scans of injected 4 MW neutral beam energy show that the beam energy has little effect on CX losses, which can be approximately estimated as 5 % of the beam power. Therefore, based on the simulation results, for EAST-NBI-1, when the line-average density \(\bar{n}_{e} = 3 \times 10^{19} /{\text{m}}^{3}\), PNBI = 2 MW and ENBI = 50 keV, the total beam losses (CX losses and shinethrough losses) are about 30 % of the beam power. These conclusions will be valuable for the experimental analysis of beam-heated discharges on EAST.

Observation of Impurity Accumulation After Hydrogen Multi-Pellet Injection in Large Helical Device

Impurity accumulation is studied for neutral beam-heated discharges after hydrogen multi-pellet injection in Large Helical Device (LHD). Iron density profiles are derived from radial profiles of EUV line emissions of FeXV-XXIV with the help of the collisional-radiative model. A peaked density profile of Fe23+ is simulated by using one-dimensional impurity transport code. The result indicates a large inward velocity of −6 m/s at the impurity accumulation phase. However, the discharge is not entirely affected by the impurity accumulation, since the concentration of iron impurity, estimated to be 3.3 × 10−5 to the electron density, is considerably small. On the other hand, a flat profile is observed for the carbon density of C6+, which is derived from the Zeff profile, indicating a small inward velocity of −1 m/s. These results suggest atomic number dependence in the impurity accumulation of LHD, which is similar to the tokamak result.

Ion energy measurements in the scrape-off layer of MAST using a Retarding Field Analyzer

We present first ion energy measurements obtained in the scrape-off layer (SOL) of the MAST tokamak using a Retarding Field Analyzer (RFA). Results from two sets of experiments are reported. First, ion temperature (Ti) profiles were measured in L-mode discharges. Ti is lower than 20eV and larger than the electron temperature with a ratio in the range 1–2.6, even in beam-heated discharges. This is consistent with a relatively high degree of ion–electron thermal coupling. Second, the RFA was used to estimate the ion energy in ELMs. Two different scenarios were studied and gave very different results. In the first one, a type-I ELMy H-mode, ions with energies exceeding 500eV were found as far as 20cm away from the separatrix. More statistics would be needed to calculate Ti in ELM filaments. In the second one, featuring type-III ELMs, no ions with energy larger than 200eV were detected 10cm outside the separatrix.

Energy and pitch angle-resolved measurements of escaping helically trapped energetic ions at the small major radius side of the compact helical system

We have developed and installed a new, second escaping fast ion probe for the small major radius side of the compact helical system. This is a Tokamak Fusion Test Reactor type scintillator-based probe and is intended to detect unconfined helically trapped fast ions whose orbits largely deviate from magnetic flux surfaces. We observed a localized light spot on the scintillator screen in neutral beam-heated discharges and it was confirmed to be a true fast ion signal. The analysis suggests that the probe detects partially thermalized, pitch angle scattered beam ions.

A mechanism for beam-driven excitation of ion cyclotron harmonic waves in the Tokamak Fusion Test Reactor

A mechanism is proposed for the excitation of waves at harmonics of the injected ion cyclotron frequencies in neutral beam-heated discharges in the Tokamak Fusion Test Reactor (TFTR) [Proceedings of the 17th European Conference on Controlled Fusion and Plasma Heating (European Physical Society, Petit-Lancy, Switzerland, 1990), p. 1540]. Such waves are observed to originate from the outer midplane edge of the plasma. It is shown that ion cyclotron harmonic waves can be destabilized by a low concentration of sub-Alfvénic deuterium or tritium beam ions, provided these ions have a narrow distribution of speeds parallel to the magnetic field. Such a distribution is likely to occur in the edge plasma, close to the point of beam injection. The predicted instability gives rise to wave emission at propagation angles lying almost perpendicular to the field. In contrast to the magnetoacoustic cyclotron instability proposed as an excitation mechanism for fusion-product-driven ion cyclotron emission in the Joint European Torus (JET) [Phys. Plasmas 1, 1918 (1994)], the instability proposed here does not involve resonant fast Alfvén and ion Bernstein waves, and can be driven by sub-Alfvénic energetic ions. It is concluded that the observed emission from TFTR can be driven by beam ions.

Observation of a localized transition from edge to core density turbulence in the TFTR tokamak.

A localized transition zone is observed between turbulent, long-wavelength density fluctuations (${\mathit{k}}_{\mathrm{\ensuremath{\perp}}}$2 ${\mathrm{cm}}^{\mathrm{\ensuremath{-}}1}$) in the core and edge of beam-heated discharges in TFTR. The core turbulence is unimodal while the edge has two counterpropagating modes, both of which are spectrally distinct from the core fluctuations. The fluctuation amplitude in the core scales with the global energy confinement time while in the edge it does not. These observations suggest that distinct modes are responsible for turbulence in the two regions and that only the core mode is directly related to global confinement.

Deuterium recycling, confinement and limiter flux in TFTR

The neutrals code DEGAS and the transport code SNAP were used to model recycling during steady state phases of Ohmic and neutral beam-heated discharges in TFTR. The flux of deuterium from the inner limiter is calculated to be 15–45 times the total D α emission rate, with ≈ 2/3 of the deuterium flux resulting from D + flow in the scrape off to the limiter, and ≈ 1/3 resulting from D° scattering off the limiter. The total D + ionization rate in the core plasma is 6–24 times the total D α emission rate, and is larger than the total neutral beam fueling rate. The D α emission rate, limiter flux, and core ionization rates scale approximately as the square of the volume averaged electron density.

Instrumental aspects of extraordinary mode scattering on TFTR

Initial results from extraordinary mode microwave scattering on TFTR show that the magnitude of density fluctuations approximately follows the mixing length prediction in both beam-heated and ohmic discharges and that the most important part of the k spectrum falls below k⊥ρs=0.5 at a scale unresolved in these measurements. These results are consistent with most scattering measurements done in the ordinary mode on smaller devices. We observe only electron diamagnetic spectral shifts in ohmic discharges while beam-heated discharges display a shift which depends on the sign of the toroidal flow velocity. The spatial resolution of the system inferred from ray tracing is consistent with the measured toroidal velocity profiles and measured spectral shapes. There are some limitations on our measurements caused by a relatively large ‘‘ghost’’ component in the low-frequency spectrum which cannot be identified with simple scattering from the crossing region of the incident and scattering antenna patterns. This component has some of the average features of scattering in the central region but does not have the frequency offset expected for drift waves and does not exhibit a Doppler shift when the plasma is rotating. We show that this ghost feature can be explained by non-Gaussian wings on the transmitting and receiving antennas which have a response to the high fluctuation levels at low values of k.

Design of the TFTR rf probe array (abstract)

A new diagnostic, composed of an array of high-frequency magnetic probes, has been installed in TFTR for the study of waves driven by the ICRF heating antennas and waves generated spontaneously during ohmic and beam-heated discharges. The former is of interest because of the possibility of surface wave generation by the heating antennas, and the latter because emission spectra may be useful as a beam ion or fusion product diagnostic. The array consists of seven fixed probes positioned at various toroidal and poloidal locations in order to obtain information about the corresponding wavenumbers in the edge region. Each probe consists of two orthogonal loops oriented to measure Bφ and Bθ. The array utilizes single-turn loops with areas of 12 cm2 and carefully selected signal processing elements to achieve a usable bandwidth of 1–500 MHz. Reduction of electrostatic pickup is accomplished by differentially combining signals from both ends of a loop using a wideband hybrid junction and by using an alumina shield to reduce capacitive coupling. The system permits simultaneous observation of both the magnetic and electrostatic components of the probe signal, allowing verification of the rejection ratio in a particular experiment. Details of the diagnostic design and experimental plans are discussed. This work was supported by U. S. DOE Contract No. DE-AC02-76-CHO-3073.

Impurity transport in ohmic and neutral-beam-heated TFTR discharges

Impurity transport has been studied in ohmic and neutral-beam-heated TFTR discharges by observation and numerical modeling of time evolutions of emissions from germanium injected using the laser-blowoff method. VUV spectral lines, the total radiated power, and soft X-ray intensities are compared with code predictions in which the impurity flux is assumed to have diffusive and convective terms. In plasmas with I p = 1.4 MA and n e ≈ 2.5 × 10 13 cm −3 , no difference in transport between ohmic and beam-heated discharges is observed. At I p = 0.7 MA and n e ≈ 1.0 × 10 13 cm −3 , a three-fold increase in the diffusion coefficient is found during beam heating.

The effect of the poloidal divertor neutralizer geometry on the scrape-off plasma temperature and density

The poloidal magnetic divertor operating in the cool-high density plasma mode represents a good solution to the tokamak first wall erosion and impurity transport problems. It also, in general, improves the energy confinement time in the beam-heated discharges. In the existing experiments, the neutralizer plates are oriented close to normal to the flux surfaces in the poloidal plane, but the planned high-power experiments require that the plate be sharply angled to reduce the power flux density. We have studied the effect of changing the plate angle on the radial plasma temperature and density profiles at the plate. We find that some plate orientations shift the electron flux profile relative to the power flux profile, resulting in unfavorable temperature profiles. The question of existence of a steady-state solution was also addressed. For a steady state to exist, the radial neutral fluxes have to exactly cancel the radial diffusive ion fluxes. Depending on the diffusion coefficients, the plasma density may have to increase with radius to satisfy this condition. The implications of these conclusions for the INTOR, and CIT experiments is also discussed.

Weight change measurements of erosion/deposition at beryllium limiter tiles in the Impurity Study Experiment-B

The weight changes of Be tiles which functioned as a rail limiter in ISX-B for more than 3500 beam-heated discharges have been determined. The net weight loss for the limiter was 2.0 g, with the central tiles losing a total of 3.2 g, and the inboard tiles gaining 1.2 g. The weight loss is attributed primarily to the release of Be droplets as a result of limiter surface melting. The weight gains resulted from an inward flow of molten material along the limiter surface. The results indicate high erosion (melt loss) with incomplete and nonuniform redeposition (melt flow) of limiter material during periods of limiter melting.

Studies of diverted plasmas and high density disruption in DITE

The latest analysis of results from the Bundle Divertor and prospects for application to a tokamak reactor are presented. The upgraded version of the machine, DITE IB is described briefly. Initial results of a programme to study the causes of disruption at high density in ohmic and beam-heated discharges are also described.

Energy transport of beam-heated high-temperature and high-density plasmas in Doublet-III

The empirical scaling of the electron thermal diffusivity, χe, is investigated for more than 100 beam-heated discharges. These discharges include two major features: 1) high-temperature and high-density plasma (Te(0)≈Ti(0)≈2.5–3 keV at n̄e≈(5–7) × 1013 cm−3, PNBI ⪅ 4 MW), which will be the basis for the breakeven experiments in the next-generation tokamaks, 2) three types of discharges, i.e. good and poor confinement divertor discharges and limiter discharges. – All kinds of discharges (good heating and poor heating divertor discharges, limiter discharges) have fhe same functional form in χe within ∼ 40% at 0.25 a < r < 0.65 a, where χe in the simplest expression scales as χe ∝ 1/ne. The ion thermal diffusivity, χi, is consistent with the assumption that the neoclassical χi can be applicable to our three kinds of discharges. For discharges with different heating efficiency, there is no systematic difference, in the adjustable multiplier, to the neoclassical theory by Hinton-Hazeltine for an ion collisionality of νi* = 0.02–0.5.

Measurement of χe in ISX-B beam-heated discharges by heat pulse propagation

New measurements of the electron thermal diffusivity coefficient χe for beam-heated (0–1.6 MW) discharges in the Impurity Study Experiment (ISX-B) tokamak confirm the numerical values and trends, with beam power inferred earlier from power balance considerations, and thus support the conclusion that the confinement deterioration with increasing beam power is due to enhanced electron heat conduction losses.

Energy confinement of beam-heated divertor and limiter discharges in Doublet III

Observation of the intensity of the recycling particle flux at the main plasma edge for various limiter and divertor discharges indicates that the gross energy confinement of beam-heated discharges is closely related to the intensity of the edge particle flux. In limiter discharges, the global particle confinement time and the energy confinement time τE show many similarities: 1) linear Ip dependence at Ip < 600 kA, 2) no BT dependence, and 3) deterioration against injection power. Improvement of τE by increasing Ip, for example, is associated with high temperatures at the plasma edge region accompanied by reduced particle recycling. – Divertor discharges with low particle recycling around the main plasma show better energy confinement than limiter discharges at high plasma densities. The improvement of τE is primarily originated in the reduction of heat transport at the main plasma edge region, which is associated with the reduction of recycling particle flux at the main plasma edge. Under certain operation condition, for example, excessive cold-gas puffing, the discharge shows relatively high scrape-off plasma density and strong particle recycling between the main plasma and the limiter. The energy confinement time of these discharges degrades somewhat or reduces completely to that of the limiter discharge. – In low-recycling divertor discharges, the central electron and ion temperature is proportional to the injection power, and the plasma stored energy is proportional to n̄ePabs (scales as INTOR scaling). With ≈ 4 MW beam injection, high-temperature and high-density plasmas were obtained (stored energy up to 280 kJ, Te(0) ≈ Ti(0) ≈ 2.5–3.0 keV at n̄e ≈ (6–7) × 1013 cm−3, τE* ≈ 70 ms).

Plasma edge studies using carbon resistance probes

A new experimental technique, the resistance probe, was used to study the plasma edge in the PLT and PDX tokamaks. This technique involves measuring the change in resistance of a thin carbon film due to bombardment by energetic particles escaping the plasma. The probes have been calibrated by measuring the resistance change caused by implantation of various ions at different energies. A model has been developed which can be used to determine the flux and energy of the incident particles from the measured resistance changes. For probes exposed in PDX and PLT near the wall, resistance changes were observed due to charge exchange neutrals. Larger changes were observed in the ion scrape-off region closer to the plasma. In PLT the effect of ions at the plasma edge begins to dominate the neutral flux near the radius of the ring limiter. The energy of ions at the plasma edge was estimated to be low (≲100 eV) in PDX during neutral beam-heated discharges, but higher (≳300 eV) in PLT during ion cyclotron resonance heating.

Plasma-wall interaction in ATC during high power neutral beam injection

Measurements of the elemental composition of samples of the vacuum vessel wall surface and of relative impurity influx rates into ATC during high power beam-heated discharges are combined with a computer simulation of the plasma and previous measurements of power balance and scaling laws to give a model of the main plasma-wall interactions in ATC. It is shown that plasma and beam charge exchange neutrals are the primary causes of impurity influx during neutral beam injection.