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© 2020 The Author(s). The SPARC tokamak is a critical next step towards commercial fusion energy. SPARC is designed as a high-field (T), compact (m, m), superconducting, D-T tokamak with the goal of producing fusion gain 2$]]> from a magnetically confined fusion plasma for the first time. Currently under design, SPARC will continue the high-field path of the Alcator series of tokamaks, utilizing new magnets based on rare earth barium copper oxide high-temperature superconductors to achieve high performance in a compact device. The goal of 2$]]> is achievable with conservative physics assumptions () and, with the nominal assumption of, SPARC is projected to attain and MW. SPARC will therefore constitute a unique platform for burning plasma physics research with high density (), high temperature (keV) and high power density () relevant to fusion power plants. SPARC's place in the path to commercial fusion energy, its parameters and the current status of SPARC design work are presented. This work also describes the basis for global performance projections and summarizes some of the physics analysis that is presented in greater detail in the companion articles of this collection.

Progress in technological fields such as high-temperature superconductors, additive manufacturing, and innovative materials has led to new scenarios and to a second generation of fusion reactor designs. The new Affordable Robust Compact (ARC) fusion reactor, which compared to other designs meets its goal to achieve fusion energy in a less expensive, smaller but even more powerful, faster way, has been designed at Massachusetts Institute of Technology. In order to define ARC’s role in future electricity grids, a feasibility investigation of the load-following concept has been carried out, starting on ARC’s vacuum vessel (VV), which is the component closest to the plasma. Finite element analysis models have been designed, and thermomechanical analyses have been conducted. In this framework thermal fatigue and creep remain the main issues. This study identifies and verifies a suitable temperature range for the VV coolant. Indeed, it is found to satisfy both requirements for the lifetime of the structural material and thermodynamic efficiency optimization.

Affordable Robust Compact reactor is a conceptual design for a Tokamak conceived by Massachusetts Institute of Technology (MIT) researchers. The design of this tokamak is under development and update. One of the key parameters for fusion reactor power plants is the tritium breeding ratio (TBR), which has to guarantee the tritium self-sufficiency. The tritium inventory circulating in a fusion power plant must be minimized. In the meantime, to enhance plant's economics, the amount of tritium generated and stored should be maximized, since it would be used to startup new reactors. Both of the aforementioned trends meet their best in a TBR as high as possible. In this work, ARC tritium breeding ratio is studied and optimized. Taking advantage of Monte Carlo neutron transport codes, several configurations of ARC's blanket and vacuum vessel have been analyzed in order to find the most effective one for a high TBR. The study takes into account different materials for the structure, such as Inconel718, V-15Cr-5Ti and Eurofer97. Moreover, it scans different width of coolant's channels and evaluates the effect of lithium-6 enrichment in the blanket looking for the best configuration in terms of TBR.

Nowadays, Fusion Energy is one of the most important sources under study. During the last years, different designs of fusion reactors were considered. At the MIT, an innovative design was created: ARC, the Affordable Robust Compact reactor. It takes advantage of the innovative aspects of recent progress in fusion technology, such as high temperature superconductors, that permit to decrease the dimensions of the machine, reaching at the same time high magnetic fields. Our main goal is the low-activation analysis of possible structural materials for the vacuum vessel, which is designed as a single-piece placed between the first-wall and the tank that contains the breeding blanket. Due to its position, the vacuum vessel is subject to high neutron flux, which can activate it and cause the reduction of the component lifetime and decommissioning problems. The activation analysis was done also for the liquid breeder FLiBe, compared with Lithium-Lead. Codes used for the low-activation analysis were MCNP and FISPACT-II. The first one is based on a neutronics model and for each component a certain neutron flux is evaluated. For FISPACT-II, the main input is the composition of the analyzed material, the neutron flux and the irradiation time. Results from FISPACT-II are the time behavior of specific activity, contact dose rate. To assess suitable structural materials for the vacuum vessel, low-activation properties were considered. Vanadium alloys turn out to be one of the best alternatives to the present material, Inconel-718. Finally, isotopic tailoring and elemental substitution methods were applied. Here, the composition of each alloy is analyzed and critical isotopes or elements are eliminated or reduced. After the modifications, new simulations are done, and those leading to significant improvements in the final results are highlighted.

The ARC pilot plant conceptual design study has been extended beyond its initial scope [B. N. Sorbom et al., FED 100 (2015) 378] to explore options for managing ∼525 MW of fusion power generated in a compact, high field (B0 = 9.2 T) tokamak that is approximately the size of JET (R0 = 3.3 m). Taking advantage of ARC's novel design – demountable high temperature superconductor toroidal field (TF) magnets, poloidal magnetic field coils located inside the TF, and vacuum vessel (VV) immersed in molten salt FLiBe blanket – this follow-on study has identified innovative and potentially robust power exhaust management solutions. The superconducting poloidal field coil set has been reconfigured to produce double-null plasma equilibria with a long-leg X-point target divertor geometry. This design choice is motivated by recent modeling which indicates that such configurations enhance power handling and may attain a passively-stable detachment front that stays in the divertor leg over a wide power exhaust window. A modified VV accommodates the divertor legs while retaining the original core plasma volume and TF magnet size. The molten salt FLiBe blanket adequately shields all superconductors, functions as an efficient tritium breeder, and, with augmented forced flow loops, serves as an effective single-phase, low-pressure coolant for the divertor, VV, and breeding blanket. Advanced neutron transport calculations (MCNP) indicate a tritium breeding ratio of ∼1.08. The neutron damage rate (DPA/year) of the remote divertor targets is ∼3–30 times lower than that of the first wall. The entire VV (including divertor and first wall) can tolerate high damage rates since the demountable TF magnets allow the VV to be replaced every 1–2 years as a single unit, employing a vertical maintenance scheme. A tungsten swirl tube FLiBe coolant channel design, similar in geometry to that used by ITER, is considered for the divertor heat removal and shown capable of exhausting divertor heat flux lev, DOE NNSA Stewardship Science Graduate Fellowship (No. DE-NA0002135), National Science Foundation Graduate Research Fellowship (Grant No. DGE1122374)

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 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)

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)

A study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE (Rognlien et al., J. Nucl. Mater. 196, 347, 1992). The model parameters are based on the ADX tokamak concept design (LaBombard et al., Nucl. Fusion 55, 053020, 2015). Several long-legged divertor configurations are considered, in particular the X-point target configuration proposed for ADX, and compared with a standard divertor. For otherwise identical conditions, a scan of the input power from the core plasma is performed. It is found that as the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remaining stable. At low power the detachment front eventually moves to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, the operation window for a detached divertor in the standard divertor configuration is very small, or even non-existent; under the same conditions for long-legged divertors the detached operation window is quite large, in particular for the X-point target configuration, allowing a factor of 5–10 variation in the input power. These modeling results point to possibility of stable fully detached divertor operation for a tokamak with extended divertor legs., United States. Department of Energy (Contract DE-AC52-07NA27344)

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)

Growth of tungsten nano-tendrils (“fuzz”) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The depth of the W fuzz layer (600 ± 150 nm) is consistent with an empirical growth formula from the PISCES experiment. Re-creating the C-Mod exposures as closely as possible in Pilot-PSI experiment can produce nearly-identical nano-tendril morphology and layer thickness at surface temperatures that agree with uncertainties with the C-Mod W probe temperature data. Helium concentrations in W fuzz layers are measured at 1–4 at.%, which is lower than expected for the observed sub-surface voids to be filled with several GPa of helium pressure. This possibly indicates that the void formation is not pressure driven., United States. Dept. of Energy (Award DE-SC00-02060)

Measurements of impurities in Alcator C-Mod indicate that in the pedestal region, significant poloidal asymmetries can exist in the impurity density, ion temperature, and main ion density. In light of the observation that ion temperature and electrostatic potential are not constant on a flux surface [Theiler et al., Nucl. Fusion 54, 083017 (2014)], a technique based on total pressure conservation to align profiles measured at separate poloidal locations is presented and applied. Gyrokinetic neoclassical simulations with XGCa support the observed large poloidal variations in ion temperature and density, and that the total pressure is approximately constant on a flux surface. With the updated alignment technique, the observed in-out asymmetry in impurity density is reduced from previous publishing [Churchill et al., Nucl. Fusion 53, 122002 (2013)], but remains substantial (nz,H/nz,L∼6). Candidate asymmetry drivers are explored, showing that neither non-uniform impurity sources nor localized fluctuation-driven transport are able to explain satisfactorily the impurity density asymmetry. Since impurity density asymmetries are only present in plasmas with strong electron density gradients, and radial transport timescales become comparable to parallel transport timescales in the pedestal region, it is suggested that global transport effects relating to the strong electron density gradients in the pedestal are the main driver for the pedestal in-out impurity density asymmetry., United States. Department of Energy (DE-FC02-99ER54512), United States. Department of Energy (DE-FG02-06ER54845), United States. Department of Energy (DE-FG02-86ER53223), United States. Department of Energy (DE-AC02-09CH11466)

This paper describes joint ITPA studies of the I-mode regime, which features an edge thermal barrier together with L-mode-like particle and impurity transport and no edge localized modes (ELMs). The regime has been demonstrated on the Alcator C-Mod, ASDEX Upgrade and DIII-D tokamaks, over a wide range of device parameters and pedestal conditions. Dimensionless parameters at the pedestal show overlap across devices and extend to low collisionality. When they are matched, pedestal temperature profiles are also similar. Pedestals are stable to peeling–ballooning modes, consistent with lack of ELMs. Access to I-mode is independent of heating method (neutral beam injection, ion cyclotron and/or electron cyclotron resonance heating). Normalized energy confinement H 98,y2 ≥ 1 has been achieved for a range of 3 ≤ q 95 ≤ 4.9 and scales favourably with power. Changes in turbulence in the pedestal region accompany the transition from L-mode to I-mode. The L–I threshold increases with plasma density and current, and with device size, but has a weak dependence on toroidal magnetic field B T. The upper limit of power for I-modes, which is set by I–H transitions, increases with B T and the power range is largest on Alcator C-Mod at B > 5 T. Issues for extrapolation to ITER and other future fusion devices are discussed., United States. Department of Energy (DE-FC02-99ER54512-CMOD), United States. Department of Energy (DE-SC0012469), United States. Department of Energy (DE-FC02-04ER54698), United States. Department of Energy (DE-FG02-94ER54235), United States. Department of Energy (DE-AC52-07NA27344), United States. Department of Energy (DE-AC02-09CH11466), United States. Department of Energy (DE-FG02-89ER53296), United States. Department of Energy (DE-FG02-08ER54999), United States. Department of Energy (DE-FG02-08ER54984)

This paper presents the first in situ time- and space-resolved measurements of deuterium (D) fuel retention in plasma-facing component (PFC) surfaces using Accelerator-based In-situ Materials Surveillance (AIMS) on the Alcator C-Mod tokamak. AIMS is a novel in situ materials diagnostic technique based on the spectroscopic analysis of nuclear reaction products induced in PFC surfaces using an ∼MeV beam of deuterons from a compact linear accelerator in between plasma shots. AIMS measurements of D retention on inner wall PFCs were acquired during diverted and limited plasma operations and during wall conditioning experiments. Intershot measurements demonstrate the local erosion and codeposition of boron films on PFC surfaces with a constant ratio D/B. This is consistent with previous results suggesting that D codeposition with boron is insufficient to account for the net retention observed in Alcator C-Mod. Changes in deuterium concentration during boronization, electron cyclotron and glow cleanings were also measured., United States. Department of Energy (Grant DE-FG02-94ER54235), United States. Department of Energy (DE-FC02-99ER54512)

United States. Department of Energy (DE-FC02-99ER54512)

Parametric decay instabilities (PDI) appear to be an ubiquitous feature of lower hybrid current drive (LHCD) experiments at high density. In density ramp experiments in Alcator C-Mod and other machines the onset of PDI activity has been well correlated with a decrease in current drive efficiency and production of fast electron bremsstrahlung. However whether PDI is the primary cause of the 'density limit', and if so by exactly what mechanism (beyond the obvious one of pump depletion) has not been clearly established. In order to further understand the connection, the frequency spectrum of PDI activity occurring during Alcator C-Mod LHCD experiments has been explored in detail by means of a number of RF probes distributed around the periphery of the C-Mod tokamak including a probe imbedded in the inner wall. The results show that (i) the excited spectra consists mainly of a few discrete ion cyclotron (IC) quasi-modes, which have higher growth than the ion sound branch; (ii) PDI activity can begin either at the inner or outer wall, depending on magnetic configuration; (iii) the frequencies of the IC quasi-modes correspond to the magnetic field strength close to the low-field side (LFS) or high-field side separatrix; and (iv) although PDI activity may initiate near the inner separatrix, the loss in fast electron bremsstrahlung is best correlated with the appearance of IC quasi-modes characteristic of the magnetic field strength near the LFS separatrix. These data, supported by growth rate calculations, point to the importance of the LFS scrape-off layer (SOL) density in determining PDI onset and degradation in current drive efficiency. By minimizing the SOL density it is possible to extend the core density regime over which PDI can be avoided, thus potentially maximizing the effectiveness of LHCD at high density. Increased current drive efficiency at high density has been achieved in FTU and EAST through lithium coating and special fuelling methods, and in recent C-Mod, United States. Department of Energy (DE-FC02-99ER54512)

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)

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)

This paper presents an overview of recent highlights from research on Alcator C-Mod. Significant progress has been made across all research areas over the last two years, with particular emphasis on divertor physics and power handling, plasma–material interaction studies, edge localized mode-suppressed pedestal dynamics, core transport and turbulence, and RF heating and current drive utilizing ion cyclotron and lower hybrid tools. Specific results of particular relevance to ITER include: inner wall SOL transport studies that have led, together with results from other experiments, to the change of the detailed shape of the inner wall in ITER; runaway electron studies showing that the critical electric field required for runaway generation is much higher than predicted from collisional theory; core tungsten impurity transport studies reveal that tungsten accumulation is naturally avoided in typical C-Mod conditions., United States. Department of Energy (DE-FC02-99ER54512-CMOD), United States. Department of Energy (DE-AC02-09CH11466), United States. Department of Energy (DE-FG02-96ER-54373), United States. Department of Energy (DE-FG02-94ER54235)

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 efficiency of LHCD on Alcator C-Mod drops exponentially with line average density. At reactor relevant densities (> 1 · 1020 [m[-3 superscript]]) no measurable current is driven. While a number of causes have been suggested, no specific mechanism has been shown to be responsible for the loss of current drive at high density. Fast modulation of the LH power was used to isolate and quantify the LHCD deposition within the plasma. Measurements from these plasmas provide unique evidence for determining a root cause. Modulation of LH power in steady plasmas exhibited no correlated change in the core temperature. A correlated, prompt response in the edge suggests that the loss in efficiency is related to a edge absorption mechanism. This follows previous results which found the generation of n||-independent SOL currents. Multiple Langmuir probe array measurements of the conducted heat conclude that the lost power is deposited near the last closed flux surface. The heat flux induced by LH waves onto the outer divertor is calculated. Changes in the neutral pressure, ionization and hard X-ray emission at high density highlight the importance of the active divertor in the loss of efficiency. Results of this study implicate a mechanism which may occur over multiple passes, leading to power absorption near the LCFS.

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 × 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 × B shear in GYRO simulations show that E × 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 × 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.

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 × 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 × B shear in GYRO simulations show that E × 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 × 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., United States. Department of Energy (Contract No. DE-FC02-99ER54512-CMOD), United States. Department of Energy. Office of Science (Contract No. DE-AC02- 05CH11231)

New measurements of electron temperature fluctuations associated with the weakly coherent mode (WCM) during improved mode, or I-mode plasmas (Whyte et al 2010 Nucl. Fusion. 50 105005) at Alcator C-Mod (Marmar et al 2007 Fusion. Sci. Technol. 51 3261) are presented in this paper. The measurements are made with a 32-channel, high-resolution profile electron cyclotron emission radiometer. The WCM electron temperature fluctuations are localized to a 1 cm region inside the last closed flux surface. The WCM electron temperature fluctuation level is measured in several different I-mode discharges and is in the range 1% [~ over T][subscript e/T[subscript e] < 2%, which is an order of magnitude smaller than the WCM density fluctuation level. The WCM edge fluctuations observed in I-mode are believed to play a role in increasing particle transport but not energy transport in the edge of I-mode plasmas. The large difference between normalized density and electron temperature fluctuation amplitudes provides new evidence that the WCM fluctuations can separately affect energy and particle transport., United States. Dept. of Energy (DE-FC02-99-ER54512-CMOD)

Growth of tungsten nano-tendrils ('fuzz') has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The thickness of the individual nano-tendrils (50–100 nm) and the depth of the layer (600 ± 150 nm) are consistent with observations from experiments on linear plasma devices. The observation of tungsten fuzz in a tokamak may have important implications for material erosion, dust formation, divertor lifetime and tokamak operations in next-step devices., National Science Foundation (U.S.) (Award DMR-08-19762), United States. Dept. of Energy (Award DE-SC00-02060), United States. Dept. of Energy (Contract DE-FC02-99ER54512)

Intrinsic rotation has been observed in I-mode plasmas from the C-Mod tokamak, and is found to be similar to that in H mode, both in its edge origin and in the scaling with global pressure. Since both plasmas have similar edge ∇T, but completely different edge ∇n, it may be concluded that the drive of the intrinsic rotation is the edge ∇T rather than ∇P. Evidence suggests that the connection between gradients and rotation is the residual stress, and a scaling for the rotation from conversion of free energy to macroscopic flow is calculated., United States. Dept. of Energy (contract DE-FC02-99ER54512), United States. Dept. of Energy (contract DEAC02- 09CH11466), United States. Dept. of Energy (contract DE-FG02-04ER54738), Korea (South). Ministry of Education, Science and Technology (MEST) (W. C. I. program), French National Research Agency (ANR) (contract ANR-06-PLAN-0084)

Parametric dependences of the heat flux footprint on the outer divertor target plate are explored in EDA H-mode and ohmic L-mode plasmas over a wide range of parameters with attached plasma conditions. Heat flux profile shapes are found to be independent of toroidal field strength, independent of power flow along magnetic field lines and insensitive to x-point topology (single-null versus double-null). The magnitudes and widths closely follow that of the “upstream” pressure profile, which are correlated to plasma thermal energy content and plasma current. Heat flux decay lengths near the strike-point in H- and L-mode plasmas scale approximately with the inverse of plasma current, with a diminished dependence at high collisionality in L-mode. Consistent with previous studies, pressure gradients in the boundary scale with plasma current squared, holding the magnetohydrodynamic ballooning parameter approximately invariant at fixed collisionality—strong evidence that critical-gradient transport physics plays a key role in setting the power exhaust channel., United States. Dept. of Energy (Award DE-AC52-06NA25396), United States. Dept. of Energy (Coop. Agreement DE-FC02-99ER54512)

The Lower Hybrid Current Drive (LHCD) system on Alcator C-Mod is designed to drive off-axis current in the core plasma, e.g. r/a<=0.8, however several interesting observations indicate that the LH waves are depositing significant power in the Scrape Off Layer (SOL) in high density L-Mode discharges. These phenomena are observed in accessible plasmas below the so-called “density limit” attributed to Parametric Decay Instabilities as omega-->2omegalh. Parallel current densities of ~500 kA/m [superscript 2] are measured on divertor Langmuir probes located just outside the separatrix during high power LH operation above n-bare~1×10[superscript 20] m[superscript −3]. These SOL currents rise at the same density at which the line integrated hard X-ray Bremsstrahlung emission drops sharply. Ray tracing/Fokker-Planck simulations over-estimate the line integrated X-ray emissivity by a factor of ~500 at n-bare = 1.4 10[superscript 20] m[superscript −3] as compared to a factor of 3–4 at n-bare = 0.6×10[superscript 20] m. Ray tracing simulations including a realistic SOL show similar trends in X-ray emission compared with the experimental observations. ©2009 American Institute of Physics, United States. Dept. of Energy (contract DE-AC02-76CH03073), United States. Dept. of Energy (contract DE-FC02-99ER54512)

A probe has been designed, constructed, and successfully used to inject methane into the DIII-D lower divertor in a manner imitating natural release by chemical erosion. This porous plug injector (PPI) probe consists of a self-contained gas reservoir with an integrated pressure gauge and a 3 cm diameter porous surface through which gas is injected into the lower divertor of the tokamak. The probe is positioned flush with the divertor target surface by means of the divertor materials evaluation system. Two gas delivery systems were developed: in the first, gas flow is regulated by a remotely controlled microvalve and in the second by a fixed micro-orifice flow restrictor. Because of the large area of the porous surface through which gas is admitted, the injected hydrocarbon molecules see a local carbon surface (>90% carbon) similar to that seen by hydrocarbons being emitted by chemical sputtering from surrounding carbon tiles. The distributed gas source also reduces the disturbance to the local plasma while providing sufficient signal for spectroscopic detection. In situ spectroscopic measurements with the PPI in DIII-D allow the direct calibration of response for measured plasma conditions from a known influx of gas., Natural Sciences and Engineering Research Council of Canada (NSERC), United States. Dept. of Energy (Grant No. DE-FC02-04ER54698), United States. Dept. of Energy (Grant No. DE-FG02-04ER54762), United States. Dept. of Energy (Grant No. DE-AC52-07NA27344), United States. Dept. of Energy (Grant No. DE-FG02-07ER54917), United States. Dept. of Energy (Grant No. DE-AC05-00OR22725), United States. Dept. of Energy (Grant No. DE-AC04-94AL85000)

The plasma facing components, first wall, and blanket systems of future tokamak-based fusion power plants arguably represent the single greatest materials engineering challenge of all time. Indeed, the United States National Academy of Engineering has recently ranked the quest for fusion as one of the top grand challenges for engineering in the 21st century. These challenges are even more pronounced by the lack of experimental testing facilities that replicate the extreme operating environment involving simultaneous high heat and particle fluxes, large time-varying stresses, corrosive chemical environments, and large fluxes of 14-MeV peaked fusion neutrons. Fortunately, recent innovations in computational modeling techniques, increasingly powerful high-performance and massively parallel computing platforms, and improved analytical experimental characterization tools provide the means to develop self-consistent, experimentally validated models of materials performance and degradation in the fusion energy environment. This article will describe the challenges associated with modeling the performance of plasma facing component and structural materials in a fusion materials environment, the opportunities to utilize high-performance computing, and two examples of recent progress., United States. Dept. of Energy. Office of Fusion Energy Sciences (Grant DE-FG02-04GR54750), Plasma Science and Innovation Center (Grant DE-SC0002209)

High-resolution charge-exchange recombination spectroscopic measurements of B[superscript 5+] ions have enabled the first spatially resolved calculations of the radial electric field (E[subscript r]) 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) E[subscript r] 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 E[subscript r] 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 E[subscript r] 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., United States Department of Energy (Cooperative Agreement No. DE-FC02-99ER54512)