Brandon Sorbom

ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets

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 Qp ≈ 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.

20 years of research on the Alcator C-Mod tokamak

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-modes 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.

An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices

This paper presents a novel particle accelerator-based diagnostic that nondestructively measures the evolution of material surface compositions inside magnetic fusion devices. The diagnostics purpose is to contribute to an integrated understanding of plasma-material interactions in magnetic fusion, which is severely hindered by a dearth of in situ material surface diagnosis. The diagnostic aims to remotely generate isotopic concentration maps on a plasma shot-to-shot timescale that cover a large fraction of the plasma-facing surface inside of a magnetic fusion device without the need for vacuum breaks or physical access to the material surfaces. Our instrument uses a compact (~1 m), high-current (~1 milliamp) radio-frequency quadrupole accelerator to inject 0.9 MeV deuterons into the Alcator C-Mod tokamak at MIT. We control the tokamak magnetic fields--in between plasma shots--to steer the deuterons to material surfaces where the deuterons cause high-Q nuclear reactions with low-Z isotopes ~5 μm into the material. The induced neutrons and gamma rays are measured with scintillation detectors; energy spectra analysis provides quantitative reconstruction of surface compositions. An overview of the diagnostic technique, known as accelerator-based in situ materials surveillance (AIMS), and the first AIMS diagnostic on the Alcator C-Mod tokamak is given. Experimental validation is shown to demonstrate that an optimized deuteron beam is injected into the tokamak, that low-Z isotopes such as deuterium and boron can be quantified on the material surfaces, and that magnetic steering provides access to different measurement locations. The first AIMS analysis, which measures the relative change in deuterium at a single surface location at the end of the Alcator C-Mod FY2012 plasma campaign, is also presented.

Alcator C-Mod: research in support of ITER and steps beyond

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, plasmamaterial 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.

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

Abstract 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 levels of up to 12 MW/m2. Several novel, neutron tolerant diagnostics are explored for sensing power exhaust and for providing feedback control of divertor conditions over long time scales. These include measurement of Cherenkov radiation emitted in FLiBe to infer DT fusion reaction rate, measurement of divertor detachment front locations in the divertor legs with microwave interferometry, and monitoring “hotspots” on the divertor chamber walls via IR imaging through the FLiBe blanket.

Development of REBCO-Based Magnets for Plasma Physics Research

The accessible performance range for most magnetic confinement plasma physics devices expands markedly with increasing magnetic flux density. The MIT Plasma Science and Fusion Center is investigating the use of high-temperature superconductors as a low cost means to significantly enhance the performance characteristics of small to moderate scale devices. Our initial investigation emphasized the no-insulation winding technique as a means to produce highly stable dc magnets for devices operating at moderate to high magnetic flux density. We present design, manufacture, and test results for a double pancake coil wound from a 500 m length of 12 mm wide REBCO tape. The coil provides on-axis magnetic flux density within an 8 cm clear bore in excess of 0.5 T when operated in liquid nitrogen and in excess of 6 T when operated in liquid helium. The ultimate aim of the program is to develop conduction-cooled HTS coil modules that can be used for both linear and toroidal plasma devices.

Engineering upgrades to the accelerator-based in-situ materials surveillance diagnostic on Alcator C-Mod

This paper presents an overview of the engineering upgrades being made to optimize the AIMS diagnostic on the Alcator C-Mod tokamak, a novel, particle accelerator-based diagnostic that can nondestructively measure the evolution of material surface compositions inside magnetic fusion devices. Three major AIMS subsystems are presented: the RFQ deuteron accelerator; the particle detectors; and the Alcator C-Mod tokamak. The combined results of the upgrades will enable AIMS to routinely map critical plasma-material interaction quantities, such as net erosion/redeposition and fusion fuel retention, over large areas of PFC surfaces between plasma shots and after the run day.

The engineering design of ARC: A compact, highfield, fusion nuclear science facility and demonstration power plant

The affordable, robust, compact (ARC) reactor conceptual design study aims to reduce the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion pilot power plant. ARC is a 200 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 is designed to use rare earth barium copper oxide (REBCO), a type of high-temperature superconductor (HTS), for its toroidal field coils. The use of HTS technology offers many advantages over traditional superconductors when applied to tokamak designs. REBCO superconductors in particular have orders of magnitude higher critical current density than traditional superconductors such as Nb3Sn at local fields greater than 20 T, enabling much higher fields to be used in the tokamak. The large allowable temperature range (up to ~90 K) of HTS allows the use of coolants other than helium and makes possible the design of joints in the toroidal field coils. This allows the vacuum vessel to be replaced quickly, lowering first wall survivability concerns and reducing the cost and operational implications of vessel failure during the experimental phase of the reactor. External current drive for ARC is provided by two inboard (high-field side) 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 blanket operation at ~900 K with single phase fluid cooling and a high-efficiency Brayton cycle, allowing for net electricity generation when operating ARC as a pilot power plant. When coupled with a demountable compact reactor design, the immersion blanket allows the vacuum vessel to be a replaceable component, eliminating the need for complex sector maintenance. The modular design of ARC allows a single machine to initially serve as an experiment and then transition to a demonstration commercial reactor.

Design and Fabrication of a DC Feeder System of New TF Magnet Power Supply for Accelerator-Based In-Situ Materials Surveillance in Alcator C-Mod

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.