Search

The Accelerator-Based In Situ Materials Surveillance (AIMS) diagnostic was developed to perform in situ ion beam analysis (IBA) on Alcator C-Mod in August 2012 to study divertor surfaces between shots. These results were limited to studying low-Z surface properties, because the Coulomb barrier precludes nuclear reactions between high-Z elements and the ∼1 MeV AIMS deuteron beam. In order to measure the high-Z erosion, a technique using deuteron-induced gamma emission and a low-Z depth marker is being developed. To determine the depth of the marker while eliminating some uncertainty due to beam and detector parameters, the energy dependence of the ratio of two gamma yields produced from the same depth marker will be used to determine the ion beam energy loss in the surface, and thus the thickness of the high-Z surface. This paper presents the results of initial trials of using an implanted depth marker layer with a deuteron beam and the method of ratios. First tests of a lithium depth marker proved unsuccessful due to the production of conflicting gamma peaks, among other issues. However, successful trials with a boron depth marker show that it is possible to measure the depth of the marker layer with the method of gamma yield ratios., United States. Department of Energy. (grant number DE-FG02-94ER54235, cooperative agreement number DEFC02-99ER54512)

New experiments on Alcator C-Mod reveal that the favorable impurity screening characteristics of the high-field side (HFS) scrape-off layer (SOL), previously reported for single null geometries, is retained in double null configurations, despite the formation of an extremely thin SOL. In balanced double-null, nitrogen injected locally into the HFS SOL is better screened by a factor of 2.5 compared to the same injection into the low field side (LFS) SOL. This result is insensitive to plasma current and Greenwald fraction. Nitrogen injected into the HFS SOL is not as well screened (only a factor of 1.5 improvement over LFS) in unbalanced double-null discharges, when the primary divertor is in the direction of B×∇B. In this configuration, impurity ‘plume’ emission patterns indicate that an opposing E × B drift competes with the parallel impurity flow to the divertor. In balanced double-null plasmas, the dispersal pattern exhibits a dominant E × B motion. Unbalanced discharges with the primary divertor opposite the direction of B×∇B exhibit excellent HFS screening characteristics – a factor of 5 enhancement compared to LFS. These data support the idea that future tokamaks should locate all RF actuators and close-fitting wall structures on the HFS and employ near-double-null magnetic topologies, both to precisely control plasma conditions at the antenna/plasma interface and to maximally mitigate the impact of local impurity sources arising from plasma-material interactions. Keywords: Alcator C-Mod; Impurity screening; Double null; High field side scrape-off layer, United States. Department of Energy (Contract DE-FC02-99ER54512)

Advanced Plasma Material Interaction (PMI) science requires in-situ time and space-resolved measurements over a large area of Plasma Facing Component (PFC) surfaces to study fuel retention & recovery, erosion & redeposition, material mixing, etc. A novel PFC diagnostic technique Accelerator-based In-situ Materials Surveillance (AIMS) has been developed for Alcator C-Mod. At present, the AIMS covers a relatively small (35 cm) poloidal section of the inner wall PFCs at a single toroidal angle; an upgrade has been proposed which will enable nearly full poloidal (124 cm) and 40 degree toroidal PFC coverage. This paper introduces the design, analysis and fabrication of the new TF magnet power supply system for this upgrade. First, the design of the busbar system and its support structure is described, which are required to carry 15 kA current for long pulse operation of up to 25 minutes and fault condition of 400 kA for 1 second. Additional elements in the power supply system include a bidirectional crowbar, varistor protection assemblies, and a high current bus switch. Secondly, multi-physics analyses involved in the design are presented. Electro-magnetic analysis is performed to evaluate the spreading load of the two current-carrying busbars while Joule heating with thermal racheting analysis is to estimate the temperature rise in the components. Structural analysis taking into account dead weight, thermal expansion, spreading load and seismic load is performed. All analyses are completed using finite element analysis software COMSOL. Analytical calculations are included to validate the FEA results. The power supply system is ready for fabrication., United States. Department of Energy (award DE-FC02-99ER54512)

The DIONISOS experiment is used to study the impact of RF helium (He) plasma on the surface morphology of tungsten (W) at a frequency of 13.56 MHz. Helium ion energy distributions with a span of 70–75 eV, while still below the sputtering threshold result in nano-tendril bundles (NTBs) and free-standing W whiskers on surfaces at 1020 K. The NTBs are distributed intragranularly with coverage of less than 10% while reaching up to 30 µm normal to the surface for He ion fluence of 7.6 × 10²⁵m⁻² and flux density of 10²²m⁻²s⁻¹. Analysis of the NTB interior and sub-surface structure is provided through focused ion beam cross section. Keywords: Tungsten fuzz; Helium; Nano-tendril; RF sheath, United States. Department of Energy (Award DE-SC00-02060), United States. Department of Energy (Grant DE-FC02-99ER54512)

Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential (Φ[subscript P]) in RF-heated discharges on C-Mod reveals that significant Φ[subscript P] enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show Φ[subscript P] enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant Φ[subscript P] enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement., United States. Dept. of Energy (DE-FC02-99ER54512)

The affordable, robust, compact (ARC) reactor is the product of a conceptual design study aimed at reducing the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a ∼200–250 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Q[subscript p] ≈ 13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ∼63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ∼23 T peak field on coil achievable with newly available REBCO superconductor technology. External current drive is provided by two innovative inboard RF launchers using 25 MW of lower hybrid and 13.6 MW of ion cyclotron fast wave power. The resulting efficient current drive provides a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing fluorine lithium beryllium (FLiBe) molten salt. The liquid blanket is low-risk technology and provides effective neutron moderation and shielding, excellent heat removal, and a tritium breeding ratio ≥ 1.1. The large temperature range over which FLiBe is liquid permits an output blanket temperature of 900 K, single phase fluid cooling, and a high efficiency helium Brayton cycle, which allows for net electricity generation when operating ARC as a Pilot power plant., United States. Department of Energy (Grant DE-FG02-94ER54235), United States. Department of Energy (Grant DE-SC008435), United States. Department of Energy. Office of Fusion Energy Sciences (Grant DE-FC02-93ER54186), National Science Foundation (U.S.) (Grant 1122374)

The retention of fuel in plasma facing components (PFCs) is important for the viability of fusion – both in terms of economics and safety. This study shows that a single, un-mitigated, disruption can remove >30× more fuel than that retained in a single, 1s, C-Mod discharge with molybdenum and tungsten PFCs. The fuel is recovered due to heating of the near-surface (∼100 μm) during the thermal and current quench periods of the disruption. A regression analysis of full-current disruptions in a dataset of 3200 discharges leads to a scaling of fuel recovered approximately proportional to W TH 1 × W MAG 2 where WTH and WMAG are the thermal and poloidal magnetic energy inside the vessel respectively. Scaling by surface area and disruption time scales to ITER indicate 5–10 MA plasmas with low thermal energy (e.g. during current rampdown) may be ideal for removing fuel from plasma-wetted surfaces.

Launching radio frequency (RF) waves from the high field side (HFS) of a tokamak offers significant advantages over low field side (LFS) launch with respect to both wave physics and plasma material interactions (PMI). For lower hybrid (LH) waves, the higher magnetic field opens the window between wave accessibility (n∥≡ck∥/ω>1−ω2pi/ω2+ω2pe/ω2ce−−−−−−−−−−−−−−−−√+ωpe/∣∣ωce∣∣) and the condition for strong electron Landau damping (n∥∼30/Te−−−−−√ with Te in keV), allowing LH waves from the HFS to penetrate into the core of a burning plasma, while waves launched from the LFS are restricted to the periphery of the plasma. The lower n∥ of waves absorbed at higher Te yields a higher current drive efficiency as well. In the ion cyclotron range of frequencies (ICRF), HFS launch allows for direct access to the mode conversion layer where mode converted waves absorb strongly on thermal electrons and ions, thus avoiding the generation of energetic minority ion tails. The absence of turbulent heat and particle fluxes on the HFS, particularly in double null configuration, makes it the ideal location to minimize PMI damage to the antenna structure. The quiescent SOL also eliminates the need to couple LH waves across a long distance to the separatrix, as the antenna can be located close to plasma without risking damage to the structure. Improved impurity screening on the HFS will help eliminate the long-standing issues of high Z impurity accumulation with ICRF. Looking toward a fusion reactor, the HFS is the only possible location for a plasma-facing RF antenna that will survive long-term. By integrating the antenna into the blanket module it is possible to improve the tritium breeding ratio compared with an antenna occupying an equatorial port plug. Blanket modules will require remote handling of numerous cooling pipes and electrical connections, and the addition of transmission lines will not substantially increase the level of complexity. The obvious engineering challenges associated with locating antenna structures on the HFS can be overcome if HFS antennas are incorporated in the overall experimental design from the start. The Advanced Divertor and radio frequency eXperiment(ADX) will include LH and ICRF antennas located on the HFS. Compact antenna designs based on proven technologies (e.g. multi-junction and “4-way splitter” antennas) fit within the available space on the HFS of ADX. Field aligned ICRF antennas are also located on the HFS. The ADX vacuum vessel design includes dedicated space for transmission lines, pressure windows, and vacuum feedthrus for accessing the HFS wall.

Helium ion energy distributions were measured in helicon wave-coupled plasmas of the dynamics of ion implantation and sputtering of surface experiment using a retarding field energy analyzer. The shape of the energy distribution is a double-peak, characteristic of radiofrequency plasma potential modulation. The broad distribution is located within a radius of 0.8 cm, while the quartz tube of the plasma source has an inner radius of 2.2 cm. The ion energy distribution rapidly changes from a double-peak to a single peak in the radius range of 0.7-0.9 cm. The average ion energy is approximately uniform across the plasma column including the double-peak and single peak regions. The widths of the broad distribution, ΔE, in the wave-coupled mode are large compared to the time-averaged ion energy, 〈E〉. On the axis (r = 0), ΔE/ 〈E〉 ≲ 3.4, and at a radius near the edge of the plasma column (r = 2.2 cm), ΔE/ 〈E〉 ∼ 1.2. The discharge parameter space is scanned to investigate the effects of the magnetic field, input power, and chamber fill pressure on the wave-coupled mode that exhibits the sharp radial variation in the ion energy distribution., United States. Department of Energy (Award DESC00-02060), United States. Department of Energy (Award DE-FC02-99ER54512)

The weakly coherent mode (WCM) in I-mode has been studied by a six-field two-fluid model based on the Braginskii equations under the BOUT++ framework for the first time. The calculations indicate that a tokamak pedestal exhibiting a WCM is linearly unstable to drift Alfven wave (DAW) instabilities and the resistive ballooning mode. The nonlinear simulation shows promising agreement with the experimental measurements of the WCM. The shape of the density spectral and location of the spectral peak of the dominant toroidal number mode n = 20 agrees with the experimental data from reflectometry. The simulated mode propagates in electron diamagnetic direction is consistent with the results from the magnetic probes in the laboratory frame, a large ratio of particle to heat diffusivity is consistent with the distinctive experimental feature of I-mode, and the value of the simulated χeat the edge is in the range of experimental errors of χefffrom the experiment. The prediction of the WCM shows that free energy is mainly provided by the electron pressure gradient, which gives guidance for pursuing future I-mode studies.

A concept of demountable toroidal field magnets for a compact fusion reactor is discussed. The magnets generate a magnetic field of 9.2 T on axis, in a 3.3 m major radius tokamak. Subcooled YBCO conductors have a critical current density adequate to provide this large magnetic field, while operating at 20 K reduces thermodynamic cooling cost of the resistive electrical joints. Demountable magnets allow for vertical replacement and maintenance of internal components, potentially reducing cost and time of maintenance when compared to traditional sector maintenance. Preliminary measurements of contact resistance of a demountable YBCO electrical joint between are presented.

In the absence of an internal particle source, plasma turbulence will impose an intrinsic relationship between an inwards pinch and an outwards diffusion resulting in a stationary density profile. The Alcator C-mod tokamak utilizes RF heating and current drive so that fueling only occurs in the vicinity of the separatrix. Discharges that transition from L-mode to I-mode are seen to maintain a self-similar stationary density profile as measured by Thomson scattering. For discharges with negative magnetic shear, an observed rise of the safety factor in the vicinity of the magnetic axis appears to be accompanied by a decrease of electron density, qualitatively consistent with the theoretical expectations. © 2012 American Institute of Physics., United States. Department of Energy. Office of Fusion Energy Sciences

The I-mode regime, routinely observed on the Alcator C-Mod tokamak, is characterized by an edge energy transport barrier without an accompanying particle barrier and with broadband instabilities, known as weakly coherent modes (WCM), believed to regulate particle transport at the edge. Recent experiments on the DIII-D tokamak exhibit I-mode characteristics in various physical quantities. These DIII-D plasmas evolve over long periods, lasting several energy confinement times, during which the edge electron temperature slowly evolves towards an H-mode-like profile, while maintaining a typical L-mode edge density profile. During these periods, referred to as I-mode phases, the radial electric field at the edge also gradually reaches values typically observed in H-mode. Density fluctuations measured with the phase contrast imaging diagnostic during I-mode phases exhibit three features typically observed in H-mode on DIII-D, although they develop progressively with time and without a sharp transition: the intensity of the fluctuations is reduced; the frequency spectrum is broadened and becomes non-monotonic; two dimensional space-time spectra appear to approach those in H-mode, showing phase velocities of density fluctuations at the edge increasing to about 10 km s−1. However, in DIII-D there is no clear evidence of the WCM. Preliminary linear gyro-kinetic simulations are performed in the pedestal region with the GS2 code and its recently upgraded model collision operator that conserves particles, energy and momentum. The increased bootstrap current and flow shear generated by the temperature pedestal are shown to decrease growth rates, thus possibly generating a feedback mechanism that progressively stabilizes fluctuations., United States. Department of Energy. Office of Fusion Energy Sciences (Award DE-FG02- 94ER54235), United States. Department of Energy. Office of Fusion Energy Sciences (Award DE-FG02-94ER54084), United States. Department of Energy. Office of Fusion Energy Sciences (Award DE-FG02-08ER54984), United States. Department of Energy. Office of Fusion Energy Sciences (Award DE-FC02-04ER54698)

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included., United States. Dept. of Energy (Cooperative Agreement DE-FC02-99ER54512), United States. Dept. of Energy (Cooperative Agreement DE-FG03-94ER-54241), United States. Dept. of Energy (Cooperative Agreement DE-AC02-78ET- 51013), United States. Dept. of Energy (Cooperative Agreement DE-AC02-09CH11466), United States. Dept. of Energy (Cooperative Agreement DE-FG02-95ER54309), United States. Dept. of Energy (Cooperative Agreement DE-AC02-05CH11231), United States. Dept. of Energy (Cooperative Agreement DE-AC52-07NA27344), United States. Dept. of Energy (Cooperative Agreement DE-FG02- 97ER54392), United States. Dept. of Energy (Cooperative Agreement DE-SC00-02060)

Experimental observations of lower hybrid current drive (LHCD) at high density on the Alcator C-Mod tokamak are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops suddenly above line-averaged densities of 10[superscript 20] m[superscript −3] (ω/ω[subscript LH] ~ 3) in single null discharges with large (≥8 mm) inner gaps, well below the density limit previously observed on limited tokamaks (ω/ω[subscript LH] ~ 2). Modelling and experimental evidence suggest that the absence of LHCD driven fast electrons at high density may be due to parasitic collisional absorption in the scrape-off layer (SOL). Experiments show that the population of fast electrons produced by LHCD at high density ([bar over n][subscript e] > 10[superscript 20] m[superscript -3]) can be increased by operating with an inner gap of less than ~5 mm with the strongest non-thermal emission in inner wall limited plasmas. A change in plasma topology from single to double null produces a modest increase in non-thermal emission at high density. Increasing the electron temperature in the periphery of the plasma (0.8 > r/a > 1.0) also results in a modest increase in non-thermal electron emission above the density limit. Ray tracing/Fokker–Planck simulations of these discharges predict the observed sensitivity to plasma position when the effects of collisional absorption in the SOL are included in the model., United States. Dept. of Energy (Award DE-FC02-99ER54512), United States. Dept. of Energy (Award DE-AC02-76CH03073)

For the first time, nonlinear gyrokinetic simulations of I-mode plasmas are performed and compared with experiment. I-mode is a high confinement regime, featuring energy confinement similar to H-mode, but without enhanced particle and impurity particle confinement [D. G. Whyte et al., Nucl. Fusion 50, 105005 (2010)]. As a consequence of the separation between heat and particle transport, I-mode exhibits several favorable characteristics compared to H-mode. The nonlinear gyrokinetic code GYRO [J. Candy and R. E. Waltz, J Comput. Phys. 186, 545 (2003)] is used to explore the effects of E x B shear and profile stiffness in I-mode and compare with L-mode. The nonlinear GYRO simulations show that I-mode core ion temperature and electron temperature profiles are more stiff than L-mode core plasmas. Scans of the input E x B shear in GYRO simulations show that E x B shearing of turbulence is a stronger effect in the core of I-mode than L-mode. The nonlinear simulations match the observed reductions in long wavelength density fluctuation levels across the L-I transition but underestimate the reduction of long wavelength electron temperature fluctuation levels. The comparisons between experiment and gyrokinetic simulations for I-mode suggest that increased E x B shearing of turbulence combined with increased profile stiffness are responsible for the reductions in core turbulence observed in the experiment, and that I-mode resembles H-mode plasmas more than L-mode plasmas with regards to marginal stability and temperature profile stiffness. (C) 2015 AIP Publishing LLC.