Search

The bulk irradiation of materials with 10-30 MeV protons promises to advance the study of radiation damage for fission and fusion power plants. Intermediate energy proton beams can now be dedicated to materials irradiation within university-scale laboratories. This paper describes the first such facility, with an Ionetix ION-12SC cyclotron producing 12 MeV proton beams. Samples are mm-scale tensile specimens with thicknesses up to 300 um, mounted to a cooled beam target with control over temperature. A specialized tensile tester for radioactive specimens at high temperature (500+ {\deg}C) and/or vacuum represents the conditions in fission and fusion systems, while a digital image correlation system remotely measures strain. Overall, the facility provides university-scale irradiation and testing capability with intermediate energy protons to complement traditional in-core fission reactor and micro-scale ion irradiation. This facility demonstrates that bulk proton irradiation is a scalable and effective approach for nuclear materials research, down-selection, and qualification., Comment: Submitted to NIM B journal

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.

Advancements in high temperature superconducting technology have opened a path toward high-field, compact fusion devices. This new parameter space introduces both opportunities and challenges for diagnosis of the plasma. This paper presents a physics review of a neutron diagnostic suite for a SPARC-like tokamak [Greenwald et al., 2018, https://doi.org/10.7910/DVN/OYYBNU ]. A notional neutronics model was constructed using plasma parameters from a conceptual device, called the MQ1 (Mission Q ≥ 1) tokamak. The suite includes time-resolved micro-fission chamber (MFC) neutron flux monitors, energy-resolved radial and tangential magnetic proton recoil (MPR) neutron spectrometers, and a neutron camera system (radial and off-vertical) for spatially-resolved measurements of neutron emissivity. Geometries of the tokamak, neutron source, and diagnostics were modeled in the Monte Carlo N-Particle transport code MCNP6 to simulate expected signal and background levels of particle fluxes and energy spectra. From these, measurements of fusion power, neutron flux and fluence are feasible by the MFCs, and the number of independent measurements required for 95% confidence of a fusion gain Q ≥ 1 is assessed. The MPR spectrometer is found to consistently overpredict the ion temperature and also have a 1000× improved detection of alpha knock-on neutrons compared to previous experiments. The deuterium-tritium fuel density ratio, however, is measurable in this setup only for trace levels of tritium, with an upper limit of nT/nD ≈ 6%, motivating further diagnostic exploration. Finally, modeling suggests that in order to adequately measure the self-heating profile, the neutron camera system will require energy and pulse-shape discrimination to suppress otherwise overwhelming fluxes of low energy neutrons and gamma radiation.

High-temperature superconductors (HTS) promise to revolutionize high-power applications like wind generators, DC power cables, particle accelerators, and fusion energy devices. A practical HTS cable must not degrade under severe mechanical, electrical, and thermal conditions; have simple, low-resistance, and manufacturable electrical joints; high thermal stability; and rapid detection of thermal runaway quench events. We have designed and experimentally qualified a vacuum pressure impregnated, insulated, partially transposed, extruded, and roll-formed (VIPER) cable that simultaneously satisfies all of these requirements for the first time. VIPER cable critical currents are stable over thousands of mechanical cycles at extreme electromechanical force levels, multiple cryogenic thermal cycles, and dozens of quench-like transient events. Electrical joints between VIPER cables are simple, robust, and demountable. Two independent, integrated fiber-optic quench detectors outperform standard quench detection approaches. VIPER cable represents a key milestone in next-step energy generation and transmission technologies and in the maturity of HTS as a technology.

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. ©2018, DOE NNSA Stewardship Science Graduate Fellowship (No. DE-NA0002135), National Science Foundation Graduate Research Fellowship (Grant No. DGE1122374)

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)

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 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.

New results on the I-mode regime of operation on the Alcator C-Mod tokamak are reported. This ELM-free regime features high energy confinement and a steep temperature pedestal, while particle confinement remains at L-mode levels, giving stationary density and avoiding impurity accumulation. I-mode has now been obtained over nearly all of the magnetic fields and currents possible in this high field tokamak (I-p 0.55-1.7 MA, B-T 2.8-8 T) using a configuration with B x del B drift away from the X-point. Results at 8 T confirm that the L-I power threshold varies only weakly with B-T, and that the power range for I-mode increases with B-T; no 8 T discharges transitioned to H-mode. Parameter dependences of energy confinement are investigated. Core transport simulations are giving insight into the observed turbulence reduction, profile stiffness and confinement improvement. Pedestal models explain the observed stability to ELMs, and can simulate the observed weakly coherent mode. Conditions for I-H transitions have complex dependences on density as well as power. I-modes have now been maintained in near-DN configurations, leading to improved divertor power flux sharing. Prospects for I-mode on future fusion devices such as ITER and ARC are encouraging. Further experiments on other tokamaks are needed to improve confidence in extrapolation.

This paper presents a novel particle accelerator-based diagnostic that nondestructively measures the evolution of material surface compositions inside magnetic fusion devices. The diagnostic's 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.

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)

Recent research on the Alcator C-Mod tokamak has focused on a range of scientific issues with particular emphasis on ITER needs and on detailed comparisons between experimental measurements and predictive models. Research on ICRF (ion cyclotron range of frequencies) heating emphasized the origins and mitigation of metallic impurities while work on lower hybrid current drive experiments have focused on linear and nonlinear wave interactions that limit efficiency at high densities in regimes with low single pass absorption. Experiments in core turbulence and transport focused on quantitative, multi-field comparisons between nonlinear gyro-kinetics simulations and experimental measurements of profiles, fluxes and fluctuations. Experiments into self-generated rotation observed spontaneous flow reversal at a critical density identical to the transition density between linear ohmic confinement and saturated ohmic confinement regimes. H-mode studies have measured pedestal widths consistent with kinetic-ballooning-mode-like instabilities, while the pedestal heights quantitatively match the EPED code predictions. Experiments with I-mode have increased the operating window for this promising edge-localized-mode-free regime. Extrapolation of I-mode to ITER suggests that the fusion gain Q ~ 10 could be possible in ITER. Investigations into the physics and scaling of the power exhaust channel width in attached enhanced D-alpha H-mode and L-mode plasma showed a direct connection between the midplane pressure-folding length and the outer divertor target footprint. The width was found to scale inversely with IP, while being independent of conducted power, BT or q95 and insensitive to the scrape-off layer connection length - a behaviour that suggests critical-gradient physics sets both pressure and heat-flux profiles.

With the first tokamak designed for full nuclear operation now well into final assembly (ITER), and a major new research tokamak starting commissioning (JT60SA), nuclear fusion is becoming a mainstream potential energy source for the future. A critical part of the viability of magnetic confinement for fusion is superconductor technology. The experience gained and lessons learned in the application of this technology to ITER and JT60SA, together with new and improved superconducting materials, is opening multiple routes to commercial fusion reactors. The objective of this roadmap is, through a series of short articles, to outline some of these routes and the materials/technologies that go with them.