Search

The benefits of operating fusion devices, such as tokamaks and stellarators, at high fields make high-temperature superconducting magnets necessary to realize a compact fusion power system. Superconducting stellarators, such as W7-X, have used standard low-temperature superconductor technology niobium-titanium. ARPA-E has recently funded a two-year project led by the startup Type One Energy and involving the Fusion Technology Institute at the University of Wisconsin-Madison, the Plasma Science to design and fabricate the first non-planar high-temperature superconductor (HTS) rare-earth barium copper oxide (REBCO) coil for a high-field stellarator based on the SPARC tokamak's VIPER cable concept. The design consists of a 1.5-turn non-planar REBCO coil supported by a pair of 3D printed stainless steel radial plates. The ultimate goals of the project are to determine if commercial REBCO tapes and additive manufacturing can be used to fabricate high field ( ⩾ 10 T ) non-planar coils with tight bending radii ( ≃ 100 m m ) and with a degradation of the critical current smaller than 20% with respect to the expected performance. In this work we present numerical analysis for non-planar coils (critical current, magnetic field map, Lorentz forces and quench aspects) at the operating conditions of 77 K and 20 K and the fabrication and testing in liquid nitrogen (77 K) of the first two non-planar demonstrators for stellarators based on a VIPER cable. The first demonstrator is a short NO n-planar V IP E R cab L e (cable demonstrator at which we will refer to as NOVEL) equipped with 100 HTS REBCO tapes and with a critical current of 5700 A at 77 K (self-field); the second is a M ult I ple turns (1.5-turns) NO n-pl AN ar coil (coil demonstrator at which we will refer to as MINOAN) equipped with 30 HTS tapes and with a critical current of 2100 A at 77 K (self-field). Both demonstrators were tested at 77 K (liquid nitrogen bath) and the results showed that—even after being bent into non-planar shapes with bend radii ≃ 100 m m —the degradation of the critical current I c was within 15%, meeting the expected goals of the project. [ABSTRACT FROM AUTHOR]

Asymmetric halo currents (HCs) can exert large net forces on the vacuum vessel and other components during disruptions on tokamaks. The displacements caused by these forces can then be amplified if these asymmetric forces rotate at frequencies resonant with the vessel. This paper reports on the investigation of a recently proposed scaling law for the disruption HC rotation frequency [Saperstein et al., "Halo current rotation scaling in post-disruption plasmas," Nucl. Fusion 62, 026044 (2022)] that combines measurements on Alcator C-Mod with those on HBT-EP. We find that a new non-circular version of the scaling law [ ⟨ f rot ⟩ m / ⟨ m ⟩ ∝ 1 B T (S / π) ] takes into consideration the dependence of frot on the poloidal structure of the MHD instability (m) driving the asymmetry and describes the disruption-averaged rotation frequency on C-Mod. Disruption rotation is also found to be insensitive to the vertical position and impurity content of the plasma at the onset of the disruption. However, a stagnation in the time evolution of frot is occasionally observed. Observations are consistent with the dominance of poloidal rotation during the disruption, which is motivated by the poloidal drift nature of the scaling law. [ABSTRACT FROM AUTHOR]

A new pseudolocal tomography algorithm is developed for soft X-ray(SXR) imaging measurements of the turbulent electron temperature fluctuations (δ Te) in tokamaks and stellarators. The algorithm overcomes the constraints of limited viewing ports on the vessel wall (viewing angle) and limited number of lines of sight (LOS). This is accomplished by increasing the number of LOS locally in a region of interest. Numerical modeling demonstrates that the wavenumber spectrum of the turbulence can be reliably reconstructed, with an acceptable number of viewing angles and LOS and suitable low SNR detectors. We conclude that a SXR imaging diagnostic for measurements of turbulent δ Te using a pseudolocal reconstruction algorithm is feasible. [ABSTRACT FROM AUTHOR]

In this study, a full convolutional neural network is trained on a large database of experimental EAST data to classify disruptive discharges and distinguish them from non-disruptive discharges. The database contains 14 diagnostic parameters from the ∼104 discharges (disruptive and non-disruptive). The test set contains 417 disruptive discharges and 999 non-disruptive discharges, which are used to evaluate the performance of the model. The results reveal that the true positive (TP) rate is ∼ 0.827, while the false positive (FP) rate is ∼0.067. This indicates that 72 disruptive discharges and 67 non-disruptive discharges are misclassified in the test set. The FPs are investigated in detail and are found to emerge due to some subtle disturbances in the signals, which lead to misjudgment of the model. Therefore, hundreds of non-disruptive discharges from training set, containing time slices of small disturbances, are artificially added into the training database for retraining the model. The same test set is used to assess the performance of the improved model. The TP rate of the improved model increases up to 0.875, while its FP rate decreases to 0.061. Overall, the proposed data-driven predicted model exhibits immense potential for application in long pulse fusion devices such as ITER. [ABSTRACT FROM AUTHOR]

The general characteristics of disruption halo currents, including current flowing paths between the components, are studied in the experiments of EAST tokamak with 'W-Like' graphite divertor and 'ITER-like' tungsten divertor. It is found that lower hybrid waves can efficiently reheat minor disruption plasma, this can drive plasma current for tens, or even hundreds, of milliseconds. The reheated plasma can lead to lower loop voltage and of course lower halo current. Additionally, it is confirmed that helium can not rapidly cool down plasmas with high thermal energy, but Argon can shut down plasma rapidly even with a little amount of gas. Until now no runaway electron plateau has been observed both in Argon and Helium mitigation experiments. The EAST disruption database shows that: (1) Larger loop voltage is driven in vertical displacement event disruptions than in major disruptions. (2) Halo currents decrease with the decrease of vertical displacement. (3) The boundary of is 0.72, less than the limit shown in IDDB. The experimental results described here can help to improve the design of the planned upgrade of the EAST lower divertor, and provide physics information for the ITER divertor. [ABSTRACT FROM AUTHOR]

Chinese Academy of Sciences Institute of Plasma Physics and Massachusetts Institute of Technology Plasma Science and Fusion Center have been collaborating on experimental advanced superconducting tokamak (EAST). Presented in this paper are RF, disruption, and thermal stress analyses of EAST antennas. Analyzed are I port four-strap and B port $2 \times 2$ strap antennas, which are currently installed on EAST. As for RF analysis, scattering parameters are checked to make sure that the antennas are loaded, and then electric field parallel to magnetic field are checked to find out if they are below the permissible level. As for disruption analysis, mechanical stresses for both straps with the support box and Faraday screen are obtained. As for thermal stress analysis, temperature and thermal stress for a typical strap and a Faraday tube were presented. All analyses were performed by COMSOL commercial finite-element analysis software package version 5 or 5.2. [ABSTRACT FROM AUTHOR]

The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well. [ABSTRACT FROM AUTHOR]

Recent results from an ITPA joint experiment to study the onset, growth, and decay of relativistic electrons (REs) indicate that loss mechanisms other than collisional damping may play a dominant role in the dynamics of the RE population, even during the quiescent Ip flattop. Understanding the physics of RE growth and mitigation is motivated by the theoretical prediction that disruptions of full-current (15 MA) ITER discharges could generate up to 10 MA of REs with 10-20 MeV energies. The ITPA MHD group is conducting a joint experiment to measure the RE detection threshold conditions on a number of tokamaks under quasi-steady-state conditions in which Vloop, ne, and REs can be well-diagnosed and compared to collisional theory. Data from DIII-D, C-Mod, FTU, KSTAR, and TEXTOR have been obtained so far, and the consensus to date is that the threshold E-field is significantly higher than predicted by relativistic collisional theory, or conversely, the density required to damp REs is significantly less than predicted, which could have significant implications for RE mitigation on ITER. [ABSTRACT FROM AUTHOR]

Application of lower hybrid range of frequencies (LHRF) waves can induce both co- and counter-current directed changes in toroidal rotation in Alcator C-Mod plasmas, depending on the target plasma current, electron density, confinement regime and magnetic shear. For ohmic L-mode discharges with good core LH wave absorption, and significant current drive at a fixed LH power near 0.8MW, the interior (r/a < 0.5) rotation increments (on a time scale of order the current relaxation time) in the counter-current direction if ne(1020 m-3) > q95/11.5, and in the co-current direction if ne(1020 m-3) < q95/11.5. All discharges with co-current rotation changes have q0 > 1, indicating a good correlation with driven current fraction, unifying the results observed on various tokamaks. For high density (ne ≥ 1.2 × 1020 m-3) L-mode target discharges, where core LH wave absorption is low, the rotation change is in the co-current direction, but evolves on a shorter momentum transport time scale, and is seen across the entire spatial profile. For H-mode target plasmas, both co- and counter-current direction increments have been observed with LHRF. The H-mode co-rotation is correlated with the pedestal temperature gradient, which itself is enhanced by the LH waves absorbed in the plasma periphery. The H-mode counter-rotation increment, a flattening of the peaked velocity profile in the core, is consistent with a reduction in the momentum pinch correlated with a steepening of the core density profile. Most of these rotation changes must be due to indirect transport effects of LH waves on various parameters, which modify the momentum flux. [ABSTRACT FROM AUTHOR]

A novel integrated charge exchange recombination spectroscopy (CXRS)/beam emission spectroscopy (BES) system is proposed for C-Mod, in which both measurements are taken from a shared viewing geometry. The supplementary BES system serves to quantify local beam densities and supplants the common calculation of beam attenuation. The new system employs two optical viewing arrays, 20 poloidal and 22 toroidal channels. A dichroic filter splits the light between two spectrometers operating at different wavelengths for impurity ion and beam neutrals emission. In this arrangement, the impurity density is inferred from the electron density, measured BES and CXRS spectral radiances, and atomic emission rates. [ABSTRACT FROM AUTHOR]

High-resolution charge-exchange recombination spectroscopic measurements of B5+ ions have enabled the first spatially resolved calculations of the radial electric field (Er) in the Alcator C-Mod pedestal region [E. S. Marmar, Fusion Sci. Technol. 51, 261 (2006)]. These observations offer new challenges for theory and simulation and provide for important comparisons with other devices. Qualitatively, the field structure observed on C-Mod is similar to that on other tokamaks. However, the narrow high-confinement mode (H-mode) Er well widths (5 mm) observed on C-Mod suggest a scaling with machine size, while the observed depths (up to 300 kV/m) are unprecedented. Due to the strong ion-electron thermal coupling in the C-Mod pedestal, it is possible to infer information about the main ion population in this region. The results indicate that in H-mode the main ion pressure gradient is the dominant contributor to the Er well and that the main ions have significant edge flow. C-Mod H-mode data show a clear correlation between deeper Er wells, higher confinement plasmas, and higher electron temperature pedestal heights. However, improved L-mode (I-mode) plasmas exhibit energy confinement equivalent to that observed in similar H-mode discharges, but with significantly shallower Er wells. I-mode plasmas are characterized by H-mode-like energy barriers, but with L-mode-like particle barriers. The decoupling of energy and particle barrier formation makes the I-mode an interesting regime for fusion research and provides for a low collisionality pedestal without edge localized modes. [ABSTRACT FROM AUTHOR]

This diagnostic measures temperature, density, and rotation for the fully stripped boron ion between the pedestal top and the plasma core with resolution consistent with the profile gradients. The diagnostic neutral beam used for the measurements generates a 50 keV, 6 A hydrogen beam. The optical systems provide views in both poloidal and toroidal directions. The imaging spectrometer is optimized to simultaneously accept 45 views as input with minimum cross-talk. In situ calibration techniques are applied for spatial location, spectral intensity, and wavelength. In the analysis, measured spectra are fitted to a model constructed from a detailed description of the emission physics. Methods for removal of interfering spectra are included. Applications include impurity and thermal transport. [ABSTRACT FROM AUTHOR]

Disruption mitigation experiments using massive gas injection (MGI) on Alcator C-Mod [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] and DIII-D [Luxon and Davis, Fusion Technol. 8, 441 (1985)] have shown that magnetohydrodynamics (MHD) plays an important role. The three-dimensional MHD code NIMROD [Sovinec et al., J. Comput. Phys. 195, 355 (2004)] has been extended to include atomic physics taken from the KPRAD code to perform simulations of MGI. Considerable benchmarking of the code has been done against Alcator C-Mod for neon and helium gas jet experiments. The code successfully captures the qualitative sequence of events observed in MGI experiments up to the end of the thermal quench. Neon jet simulations also show quantitative agreement with the experimental thermal quench onset time. For helium gas jets, we show that a small percent boron density can significantly alter the results even in the presence of a helium jet with three orders of magnitude higher density. The thermal quench onset time is considerably overpredicted unless boron radiation is included. A DIII-D helium jet simulation shows a faster rise time for total radiated power than the experiment, but comparable amplitude. Similar to the important role of boron in C-Mod, carbon radiation is a significant factor in DIII-D helium jet simulations and experiments. [ABSTRACT FROM AUTHOR]

Charge-exchange recombination spectroscopy can be used as a diagnostic for local densities of fully stripped impurities. For example, impurity profiles are valuable for transport studies. A major impediment to analysis is the availability of cross sections which relate the observed spectroscopic emission to the impurity densities. In this work, this impediment is removed for boron ions. Emission cross sections for B+4 states excited by charge exchange between beams of neutrals and fully stripped B+5 impurities in high density plasmas are computed for beams with multiple energy components. We show that even the low-energy neutral beam components make significant contributions to the emission for high-density plasmas which are commonly generated in the laboratory. Excited states as high as n=3 are important. This implies that a detailed description of the beam components including density, excited state population, and penetration is required for prediction of intensities of spectral lines excited by diagnostic neutral beams in high-density plasmas. Plasma experiments performed on the Alcator C-Mod tokamak will be described which clearly show the efficacy of our analysis method. This technique can be extended to species other than boron. [ABSTRACT FROM AUTHOR]

A four-channel compact neutral particle analyzer (CNPA) based on operating small Si diode detectors in pulse-height analysis (PHA) mode is used to measure energetic hydrogen minority ions with energies between ∼50 and 350 keV stemming from ion-cyclotron range-of-frequency heated D(H) Alcator C-Mod plasmas with both active and passive charge exchange (CX). First core minority ion distribution results from Alcator C-Mod discharges and a detailed description of the diagnostic are presented. The diagnostic employs integrated electronics and fast digitization of the shaping amplifier voltage. The digitized data are stored for postshot PHA, which removes the constraints of real-time PHA and allows for improved performance via elimination of base line shift effects and potentially relieving pileup through Gaussian fitting routines. The CNPA is insensitive to the large gamma and neutron background in Alcator C-Mod discharges but is susceptible to the plasma’s soft x-ray flux. The soft x-ray flux limits the CNPA energy resolution to ∼15–20 keV. A simple model is used to interpret the active CNPA data which permits rapid estimates of the core hydrogen minority temperatures and anisotropy with a time resolution of ∼100 ms. Hydrogenlike boron is identified as an important electron donor for the CX signal. [ABSTRACT FROM AUTHOR]

Alfvén eigenmodes sAEsd are studied to assess their stability in high density reactor relevant regimes where TI≈Te and as a diagnostic tool. Stable AEs are excited with active magnetohydrodynamics antennas in the range of the expected AE frequency. Toroidal Alfvén eigenmode (TAE) damping rates between 0.5%<&#x03B3/ω 4.5% have been observed in diverted and limited Ohmic plasmas. Unstable AEs are excited with a fast ion tail driven by H minority ion cyclotron radio frequency sICRFd heating with electron densities in the range of n¯e=0.5-2 × 1020 m-3. Energetic particle modes or TAEs have been observed to decrease in frequency and mode number with time up to a large sawtooth collapse, indicating the role fast particles play in stabilizing sawteeth. In the current rise phase, unstable modes with frequencies that increase rapidly with time are observed with magnetic pick-up coils at the wall and phase contrast imaging density fluctuation measurements in the core. Modeling of these modes constrains the calculated safety factor profile to be very flat or with slightly reversed shear. AEs are found to be more stable for an inboard than for central or outboard ICRF resonances in qualitative agreement with modeling. [ABSTRACT FROM AUTHOR]

A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B21/BT), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10-4. This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion Technol. 8, 441 (1985)], and C-Mod, with matching applied mode spectra, have been carried out to determine more definitively the field and size scalings. Locked modes on C-Mod are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement, frequently leading to disruptions. Intrinsic error fields inferred from the threshold studies are found to be consistent in amplitude and phase with a comprehensive model of the sources of field errors based on “as-built” coil and bus-work details and coil imperfections inferred from measurements using in situ magnetic diagnostics on dedicated test pulses. Use of the A-coils to largely cancel the 2/1 component of the intrinsic nonaxisymmetric field has led to expansion of the accessible operating space in C-Mod, including operation up to 2 MA plasma current at 8 T. [ABSTRACT FROM AUTHOR]

Factor of ∼2 higher power thresholds for low- to high-confinement mode transitions (L-H) with unfavorable x-point topologies in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are linked to flow boundary conditions imposed by the scrape-off layer (SOL). Ballooning-like transport drives flow along magnetic field lines from low- to high-field regions with toroidal direction dependent on upper/lower x-point balance; the toroidal rotation of the confined plasma responds, exhibiting a strong counter-current rotation when B×∇B points away from the x point. Increased auxiliary heating power (rf, no momentum input) leads to an L-H transition at approximately twice the edge electron pressure gradient when B×∇B points away. As gradients rise prior to the transition, toroidal rotation ramps toward the co-current direction; the H mode is seen when the counter-current rotation imposed by the SOL flow becomes compensated. Remarkably, L-H thresholds in lower-limited discharges are identical to lower x-point discharges; SOL flows are also found similar, suggesting a connection. [ABSTRACT FROM AUTHOR]

Nonlinear gyrokinetic simulations of trapped electron mode (TEM) turbulence, within an internal particle transport barrier, are performed and compared with experimental data. The results provide a mechanism for transport barrier control with on-axis radio frequency heating, as demonstrated in Alcator C-Mod experiments [S. J. Wukitch et al., Phys. Plasmas 9, 2149 (2002)]. Off-axis heating produces an internal particle and energy transport barrier after the transition to enhanced D ahigh confinement mode. The barrier foot reaches the half-radius, with a peak density 2.5 times the edge density. While the density profile peaks, the temperature profile remains relatively unaffected. The peaking and concomitant impurity accumulation are controlled by applying modest central heating power late in the discharge. Gyrokinetic turbulence simulations of the barrier formation phase, using the GS2 code [W. Dorland et al., Phys. Rev. Lett. 85, 5579 (2000)] show that toroidal ion temperature gradient driven modes are suppressed inside the barrier foot, but continue to dominate in the outer half-radius. As the density gradient steepens further, trapped electron modes are driven unstable. The onset of TEM turbulence produces an outflow that strongly increases with the density gradient, upon exceeding a new nonlinear critical density gradient, which significantly exceeds the linear critical density gradient. The TEM turbulent outflow ultimately balances the inward Ware pinch, leading to steady state. Moreover, the simulated turbulent particle diffusivity matches that inferred from particle balance using measured density profile data and the calculated Ware pinch. This turbulent diffusivity exhibits a strong unfavorable temperature dependence that allows control with central heating. [ABSTRACT FROM AUTHOR]

Simultaneous high resolution measurements of the soft x-ray emission at the top and outboard edge of the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] are presented. Impurity density profiles are derived from these measurements. The impurity density varies considerably on a flux surface in the high confinement mode (H-mode) pedestal region. The soft x-ray and impurity density pedestals are not on the same flux surface at the top and outboard edge. The pedestal widths are consistently larger at the outboard edge in the Enhanced D-alpha (EDA) H-mode, suggesting a ballooning-like character of the EDA quasi-coherent mode. The usual separation of time scales between radial and parallel impurity transport does not hold in the H-mode pedestal region of Alcator C-Mod. Thus, one should not necessarily expect that the impurity density be constant on a flux surface in this region. [ABSTRACT FROM AUTHOR]

Double transport barrier modes (simultaneous core and edge transport barrier) have been observed with off-axis ion cyclotron range of frequencies (ICRF) heating in the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. An internal transport barrier (ITB) is routinely produced in enhanced D[sub α]H-mode (EDA) discharges where the minority ion cyclotron resonance layer is at r/a∼|0.5| during the current flat top phase of the discharge. The density profile becomes peaked without the presence of a particle source in the plasma core and continues to peak until the increased core impurity radiation arrests the improved energy confinement, ultimately leading to a barrier collapse. With the addition of moderate (0.6 MW) central ICRF heating, the double barrier mode was maintained for as long as the ICRF power was applied. Modeling shows that the internal thermal barrier was maintained throughout the discharge. The presence of sawteeth throughout most of the ITB discharge allows sawtooth-induced heat pulse analysis to be performed. This analysis indicates that there is an abrupt radial discontinuity in the heat pulse time to peak profile when an ITB is present. Furthermore, this discontinuity appears to move into the core plasma from the edge region in about 0.2 s, several confinement times. The deduced thermal diffusivity, χ[sup hp], indicates that a barrier exists in the electron thermal transport, the barrier is limited to a narrow radial region, and the transport is unaffected outside this narrow radial extent. © 2002 American Institute of Physics. [ABSTRACT FROM AUTHOR]

The formation of internal transport barriers (ITBs) has been observed in the core region of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under a variety of conditions. The improvement in core confinement following pellet injection (pellet enhanced performance or PEP mode) has been well documented on Alcator C-Mod in the past. Recently three new ITB phenomena have been observed which require no externally applied particle or momentum input. Short lived ITBs form spontaneously following the high confinement to low confinement mode transition and are characterized by a large increase in the global neutron production (enhanced neutron or EN modes). Experiments with ion cyclotron range of frequencies power injection to the plasma off-axis on the high field side results in the central density rising abruptly and becoming peaked. The ITB formed at this time lasts for ten energy confinement times. The central toroidal rotation velocity decreases and changes sign as the density rises. Similar spontaneous ITBs have been observed in ohmically heated H-mode plasmas. All of these ITB events have strongly peaked density profiles with a minimum in the density scale length occurring near r/a = 0.5 and have improved confinement parameters in the core region of the plasma. [ABSTRACT FROM AUTHOR]

High resolution measurements on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1551 (1994)] of the transport barrier in the “Enhanced D[α]” (EDA) regime, which has increased particle transport without large edge localized modes, show steep density and temperature gradients over a region of 2-5 mm, with peak pressure gradients up to 12 MPa/m. Evolution of the pedestal at the L-H transition is consistent with a large, rapid drop in thermal conductivity across the barrier. A quasi-coherent fluctuation in density, potential, and B[sub pol], with ƒ[sub 0]∼ 50-150 kHz and k[sub θ] 4 cm[sup -l], always appears in the barrier during EDA, and drives a large particle flux. Conditions to access the steady-state EDA regime in deuterium include δ>0.35, q[sub 95]>3.5, and L-mode target density n[sub e]>l.2× 10[sup 20]m[sup -3]. A reduced q[sub 95] limit is found for hydrogen discharges. [ABSTRACT FROM AUTHOR]

A high spatial resolution CCD based one-dimensional imaging system to measure visible continuum emissivity profiles from Alcator C-Mod tokamak plasmas is described. The instrument has chordal resolution that is better than 1 mm for the edge region of the plasma, where very sharp (1 to 10 mm) gradient lengths in plasma parameters are observed after the formation of the H-mode transport barrier. Each image has up to 2048 pixels, and total spatial coverage goes from 2 cm inside of the magnetic axis to ∼4 cm outside of the last closed flux surface in the ∼22 cm horizontal minor radius plasmas. Time resolution can be varied from 0.21 ms to 4 ms; good signal to noise is achieved with 1 ms integration under typical plasma conditions. The emission over most of the plasma volume is dominated by free–free bremsstrahlung, and can be used to infer local values of the average ion charge (Z[sub eff]). Toroidally localized puffing of deuterium, nitrogen, and helium reveals that a significant contribution to the signal in the scrape-off layer at the extreme edge of the plasma can come from diatomic molecular band pseudocontinuum emission. © 2001 American Institute of Physics. [ABSTRACT FROM AUTHOR]

Presents impurity toroidal rotation and transport in Alcator C-Mod ohmic high confinement mode plasmas. Observation on rotation velocities; Similarity in characteristics of ohmic H-mode plasmas and ion cyclotron range of frequencies; Presence of very long impurity confinement times.

A soft x-ray tomography diagnostic has been built and operated on the Alcator C tokamak with the use of an 80-channel array of detectors. The detector system consists of miniature PIN photodiodes which are located around the periphery of the plasma at one toroidal location (top and bottom ports) in order to image the potoidal cross section of the plasma's soft x-ray emissivity. The compact size and inexpensive cost of these detectors allow for the large number of views required to avoid any assumptions about plasma rotation, position, or symmetry in the reconstruction algorithm. The use of logarithmic amplifiers in the electronic circuitry makes it possible to record conveniently data from a wide range of plasma operating parameters. This diagnostic has been used to study plasma position and shape, magnetohydrodynamic (MHD) phenomena, and the effects of impurities. Data from a pellet-fueled discharge exhibiting largescale MHD activity are presented. The system has recently been expanded to 128 detectors with the addition of an array on a side port in order to improve its coverage of the plasma and the resolution of the reconstructions. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Studies the characteristics of high-confinement modes in the Alcator C Mod high-field divertor tokamak. Charge exchange flux observed during the mode; Pressures of plasma with the best energy confinement; Edge electron temperature threshold for sustaining the mode; Hysteresis in the threshold power between low-high and high-low transitions.

Diagnostic neutral beams (DNBs) are used on tokamaks and stellarators for measurement of ion temperature, plasma current density, and other critical quantities. It is important to optimize the density in the energy components of the DNB for the diagnostic application. Measured component densities are also useful for interpretation of beam diagnostics and for design of new diagnostics. The spectrally resolved photon emission from the interaction of the neutral beam with a background neutral gas is commonly used for this measurement. The emission cross sections available for interpretation of the measurement are reviewed here. Simulations employ these cross sections with other data to infer the properties of the plasma sources from which the beams are extracted. The simulations are also reviewed. Empirical examples are drawn from beam emission spectra measured for beams installed at the Alcator C-Mod facility. [ABSTRACT FROM AUTHOR]

The full package of magnetics diagnostics for Alcator C-Mod is now installed inside the vacuum vessel. This consists of four poloidal arrays of pickup coils for measuring B[sub p], three Rogowski coils, a set of full flux loops, two diamagnetic loops with compensation coils, a small 2-D array of Mirnov coils to measure edge fluctuations, and some partial flux loops for detecting locked modes. All the coils have been absolutely calibrated to ± 0.5% before installation, and an in situ calibration of this system will also be carried out. Internal mounting enables fast response despite the shielding effects of skin currents in the thick vessel wall. However, this complicates the design considerably due to constraints on physical size, vacuum compatibility, temperature extremes, protection from plasma damage, etc. [ABSTRACT FROM AUTHOR]

The neutron production rate on Alcator C-Mod is expected to peak at 1 × 10[sup 16] n/s for the highest performance experimental pulses. Measurement of the total fusion yield is important for both assessing the experiment and for evaluation of radiological data. The neutron detectors must therefore be capable of accurate measurement over a wide operating regime (10[sup 11]-10[sup 16] n/s from the source). This can be achieved with a series of fission detectors (U[sup 235] and U[sup 238]), capable of operating in both a pulsed and current mode. Computer calculations optimizing the placement of these detectors and also the moderator design are detailed. Calibration of this system will be achieved using a Cf[sup 252] source which will be moved on a track continuously around the major axis. It will also be placed discretely at a number of locations in the torus. MCNP calculations will complement the calibration data. Details of the experimental setup and calibration scheme will be presented. [ABSTRACT FROM AUTHOR]

The Alcator C-Mod experiment is expected to produce neutrons at a rate of up to 1 × 10[sup 16] n/s during its highest performance operation. The global neutron detection system must provide accurate measurement of this production over the full operating regime of the experiment, 10[sup 11]-10[sup 16] n/s. This is accomplished using a series of uranium fission detectors enriched in U[sup 235] of staggered sensitivity. These are distributed in two moderator assemblies. A BF[sub 3] detector is added to each moderator to cover the lowest operational range. Preliminary calibration of the system has been accomplished using a 65.4 µg Cf[sup 252] source placed inside the Alcator C-Mod vacuum vessel. The system design and calibration is discussed herein. [ABSTRACT FROM AUTHOR]

Ion cyclotron resonance heating is the primary auxiliary heating on the Alcator C-Mod tokamak and is commonly used on other devices, and is planned for use on ITER. The RF-power density on C-Mod is above 5 MW m−3 providing for a unique opportunity to study wave–particle effects in the high RF power per particle regime. Minority heating produces a highly energetic tail in the minority distribution function which is measured using a compact neutral particle analyser. In this paper, we present the measurements of the fast-ion spectrum between 200 and 2 MeV, compiled over an entire experimental campaign. We also estimate the effective tail temperatures for the fast-ion distribution. We find that the fast-ion distribution is less energetic and less dense with increasing electron density; is more energetic with increasing plasma current; and is more dense but has no measurable change in energy with increasing RF power. Some possible explanations for these findings are discussed. [ABSTRACT FROM AUTHOR]

A scintillator-based energetic ion loss detector has been successfully commissioned on the Alcator C-Mod tokamak. This probe is located just below the outer midplane, where it captures ions of energies up to 2 MeV resulting from ion cyclotron resonance heating. After passing through a collimating aperture, ions impact different regions of the scintillator according to their gyroradius (energy) and pitch angle. The probe geometry and installation location are determined based on modeling of expected lost ions. The resulting probe is compact and resembles a standard plasma facing tile. Four separate fiber optic cables view different regions of the scintillator to provide phase space resolution. Evolving loss levels are measured during ion cyclotron resonance heating, including variation dependent upon individual antennae. [ABSTRACT FROM AUTHOR]

Two high resolution edge x-ray imaging diagnostics have been installed in the Alcator C-Mod tokamak. One array measures the radial soft x-ray emissivity profiles at the top of the plasma with 1.2 mm radial resolution, mapped along flux surfaces to the midplane, whereas the other measures the radial soft x-ray emissivity profiles at the outboard edge with 1.5 mm radial resolution mapped to the midplane. The two diagnostics measure the chord brightness profiles, which are then inverted to get soft x-ray emissivity simultaneously with a 12 μs sampling time. This allows us to determine if the soft x-ray emissivity, and therefore the fluorine density, is constant on a flux surface during steady state high confinement mode conditions, as well as during fast transient edge events, such as edge localized modes or transitions from the high confinement mode to the low confinement mode. Measurements are presented showing that the soft x-ray emissivity is not constant on a flux surface, but instead shows a large poloidal variation, contrary to what is assumed in the inversion routine. The effects of the poloidal variation on the inversion accuracy are estimated numerically. It is found that the emissivity is systematically overestimated at the top, and underestimated at the outboard edge, by less than 15%. The width of the x-ray emissivity pedestal is accurate to within 15%, and the location of the pedestal is accurate to within 1 mm. Measurements showing a poloidal propagation delay for the onset of the transition from high confinement mode (H mode) to low confinement mode (L mode) are also presented. © 2000 American Institute of Physics. [ABSTRACT FROM AUTHOR]