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Superconducting magnet, employed in conjunction with five types of superconducting cables in a nested solenoid configuration, produces total, central magnetic field strengths approaching 70 kG. The multiple coils permit maximum information on cable characteristics to be gathered from one test.
Design of magnetic mirror machines for controlled fusion research apparatus, and optimization of magnetic field coils by adiabatic confinement of largest number of particles for given cost
Split pair of superconducting magnetic field coils of zirconium alloy wire for plasma physics research
Shielding effectiveness and system mass of different magnetic field configurations - Electric properties and magnetic behavior of superconducting coils of niobium-zirconium alloy
A novel flip-coil measurement system has been developed for the National Synchrotron Light Source II (NSLS-II) at Brookhaven National Laboratory. This paper describes the design, implementation, and commissioning of the new measurement bench, highlighting its key features, including improved mechanical stability, advanced data acquisition, and enhanced reproducibility. The system enables precise characterization of field integrals and multipole components, ensuring the optimal performance of Insertion Devices (IDs) before installation in the NSLS-II storage ring. The flip-coil system incorporates an innovative approach to minimize mechanical and electrical errors, which significantly improves the reproducibility of measurements. In addition, the system features a state-of-the-art data acquisition system that enables real-time monitoring and analysis, further enhancing the efficiency and accuracy of the measurement process. Furthermore, preliminary tests have demonstrated that the new system meets the stringent requirements for magnetic field characterization of advanced insertion devices, making it an essential tool for future ID commissioning and quality assurance at NSLS-II.
The ARC pilot plant conceptual design study has been extended beyond its initial scope to explore options for managing ∼525 MW of fusion power generated in a compact, high field (B 0 = 9.2 T) tokamak that is approximately the size of JET (R 0 = 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. Here 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/m 2 . 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.
Pulsed inductive plasma accelerators are spacecraft propulsion devices in which energy is stored in a capacitor and then discharged through an inductive coil. The device is electrodeless, inducing a current s heet in a plasma located near the face of the coil. The propellant is accelerated and expelled at a high exhaust velocity (order of 10 km/s) through the interaction of the plasma current and the induced magne tic field, The Faraday Accelerator with RF-Assisted Discharge (FARAD) thruster is a type of pulsed inductive plasma accelerator in which t he plasma is preionized by a mechanism separate from that used to for m the current sheet and accelerate the gas. Employing a separate preionization mechanism allows for the formation of an inductive current s heet at much lower discharge energies and voltages than those used in previous pulsed inductive accelerators like the Pulsed Inductive Thr uster (PIT). In this paper, we present the design of a benchtop FARAD thruster with all the subsystems (mass injection, preionization, and acceleration) integrated into a single unit. Design of the thruster follows the guidelines and similarity performance parameters presented elsewhere. The system is designed to use the ringing, RF-frequency s ignal produced by a discharging Vector Inversion Generator (VIG) to p reionize the gas. The acceleration stage operates on the order of 100 J/pulse and can be driven by several different pulsed powertrains. These include a simple capacitor coupled to the system, a Bernardes and Merryman configuration, and a pulsecompression circuit that takes a temporally broad, low current pulse and transforms it into a short, h igh current pulse. A set of applied magnetic field coils are integrated into the system to guide the preionized propellant as it spreads ov er the face of the inductive acceleration coil. The coils are operate d in a pulsed mode, and the thruster can be operated without using the coils to determine if there is a performance improvement gain realiz ed when an applied field is present.
The Z-pinch configuration offers the promise of a compact fusion device owing to its simple geometry, unity beta, and absence of external magnetic field coils. Increasing the axial current compresses the plasma, resulting in a rapid rise of the fusion reaction rate. Historically, the Z pinch has been plagued by fast growing instabilities that limit plasma lifetimes. Recent progress has resulted from investigating approaches that provide stability. One approach exploits sheared-flow stabilization to produce an equilibrium Z pinch, which sustains the compressed plasma state for durations much longer than other plasma timescales. Recent experimental and simulation results encourage pursuit of the stabilized Z pinch to explore its fusion performance limits.
Stellarator power plants require a plan for the removal of the particles and the heat that are exhausted across the plasma edge. Unless a flowing liquid metal can be used to carry the helium exhaust to places where it can be removed from the plasma chamber, the particle exhaust must be magnetically diverted into pumping chambers. Studies are required to determine how magnetic features relate to the required divertor properties, how these magnetic features can be produced, and how they can be controlled. General studies are clarified and simplified by the use of the magnetic field line Hamiltonian $ψ_p(ψ,θ,φ)$ and a vector $\vec{x}(ψ,θ,φ)$ that gives the point in space associated with each point in the $(ψ,θ,φ)$ canonical coordinates, a flux and two angles. The non-resonant Fourier terms in ψp can be removed by a canonical transformation, so only resonant Fourier terms can determine the field line properties in the plasma edge and divertor. This paper discusses the important divertor properties and explains how $ψ_p(ψ,θ,φ)$ and $\vec{x}(ψ,θ,φ)$ can be obtained numerically in a special form for any stellarator magnetic field, $\vec{B}(\vec{x})$. This form holds between an arbitrary magnetic surface and the chamber walls with the non-resonant terms eliminated. Studies based on variations in the terms in such derived field-line Hamiltonians can determine what magnetic features are mathematically possible and how they could be produced and controlled by the external magnetic field coils.
Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is ${{R}_{0}}\geqslant 1.7$ m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.
The mission of the spherical tokamak NSTX-U is to investigate the physics that drives core and pedestal transport and stability at high-β and low collisionality, as part of the development of the ST concept towards a compact, low-cost ST-based Pilot Plant. NSTX- U will ultimately operate at up to 2 MA and 1 T with up to 12 MW of Neutral Beam Injection (NBI) power for 5 seconds. NSTX-U will operate in a regime where electromagnetic instabilities are expected to dominate transport, and beam-heated NSTX-U plasmas will explore a portion of energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U will also develop the physics understanding and control tools to ramp-up and sustain high performance plasmas in a fully- noninductive fashion. NSTX-U began research operations in 2016, but a failure of a divertor magnetic field coil after ten weeks of operation resulted in the suspension of operations and initiation of Recovery activities. During this period, there has been considerable work in the area of analysis, theory and modeling of data from both NSTX and NSTX-U, with a goal of understanding the underlying physics to develop predictive models that can be used for high-confidence projections for both ST and higher aspect ratio regimes. The studies have presented issues in thermal plasma transport, macrostability, energetic particlet-driven instabilities at ion-cyclotron frequencies and below, and edge and divertor physics.
Equilibria in magnetic confinement devices result from force balancing between the Lorentz force and the plasma pressure gradient. In an axisymmetric configuration like a tokamak, such an equilibrium is described by an elliptic equation for the poloidal magnetic flux, commonly known as the Grad–Shafranov equation. It is challenging to develop a scalable and accurate free-boundary Grad–Shafranov solver, since it is a fully nonlinear optimization problem that simultaneously solves for the magnetic field coil current outside the plasma to control the plasma shape. In this work, we develop a Newton-based free-boundary Grad–Shafranov solver using adaptive finite elements and preconditioning strategies. The free-boundary interaction leads to the evaluation of a domain-dependent nonlinear form of which its contribution to the Jacobian matrix is achieved through shape calculus. The optimization problem aims to minimize the distance between the plasma boundary and specified control points while satisfying two nontrivial constraints, which correspond to the nonlinear finite element discretization of the Grad–Shafranov equation and a constraint on the total plasma current involving a nonlocal coupling term. The linear system is solved by a block factorization, and AMG is called for subblock elliptic operators. The unique contributions of this work include the treatment of a global constraint, preconditioning strategies, nonlocal reformulation, and the implementation of adaptive finite elements. Furthermore, it is found that the resulting Newton solver is robust, successfully reducing the nonlinear residual to 1e-6 and lower in a small handful of iterations while addressing the challenging case to find a Taylor state equilibrium where conventional Picard-based solvers fail to converge.
A Fusion Nuclear Science Facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR approximately 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions vs. configuration studies including dependence on plasma major radius R0 for a range 1m to 2.2m are described. In particular, it is found the threshold major radius for TBR = 1 is R0 greater than or equal to 1.7m, and a smaller R0=1m ST device has TBR approximately 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A=2, R0=3m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.
Effective generation of helicon waves for high density plasma sources is determined by the dispersion relation and plasma power balance. Helicon wave plasma sources inherently require an applied magnetic field of .01-0.1 T, an antenna properly designed to couple to the helicon wave in the plasma, and an rf power source in the 10-100 s of MHz, depending on propellant choice. For a plasma thruster, particularly one with a high specific impulse (>2000 s), the physics of the discharge would also have to address the use of electron cyclotron resonance (ECR) heating and magnetic expansion. In all cases the system design includes an optimized magnetic field coil, plasma source chamber, and antenna. A preliminary analysis of such a system, calling on experimental data where applicable and calculations where required, has been initiated at Glenn Research Center. Analysis results showing the mass scaling of various components as well as thruster performance projections and their impact on thruster size are discussed.
Effective generation of helicon waves for high density plasma sources is determined by the dispersion relation and plasma power balance. Helicon wave plasma sources inherently require an applied magnetic field of .01-0.1 T, an antenna properly designed to couple to the helicon wave in the plasma, and an rf power source in the 10-100 s of MHz, depending on propellant choice. For a plasma thruster, particularly one with a high specific impulse (>2000 s), the physics of the discharge would also have to address the use of electron cyclotron resonance (ECR) heating and magnetic expansion. In all cases the system design includes an optimized magnetic field coil, plasma source chamber, and antenna. A preliminary analysis of such a system, calling on experimental data where applicable and calculations where required, has been initiated at Glenn Research Center. Analysis results showing the mass scaling of various components as well as thruster performance projections and their impact on thruster size are discussed.
The development and application of a multidimensional numerical simulation code for investigating near-field plasma processes in magnetic nozzles are presented. The code calculates the time-dependent evolution of all three spatial components of both the magnetic field and velocity in a plasma flow, and includes physical models of relevant transport phenomena. It has been applied to an investigation of the behavior of plasma flows found in high-power thrusters, employing a realistic magnetic nozzle configuration. Simulation of a channel-flow case where the flow was super-Alfvenic has demonstrated that such a flow produces adequate back-emf to significantly alter the shape of the total magnetic field, preventing the flow from curving back to the magnetic field coil in the near-field region. Results from this simulation can be insightful in predicting far-field behavior and can be used as a set of self-consistent boundary conditions for far-field simulations. Future investigations will focus on cases where the inlet flow is sub-Alfvenic and where the flow is allowed to freely expand in the radial direction once it is downstream of the coil.
Superconducting cyclotrons are increasingly employed for proton beam radiotherapy treatment (PBRT). The use of superconductivity in a cyclotron design can reduce its mass by an order of magnitude and size by a factor of 3-4 over conventional resistive magnet technology, yielding significant reduction in overall cost of the device, the accelerator vault, and its infrastructure, as well as reduced operating costs. At MIT, previous work was focused on developing a very high field (9 T at the pole face) superconducting synchrocyclotron that resulted in a highly compact device that is about an order of magnitude lighter, and much smaller in diameter than a conventional, resistive cyclotron. The results of the study reported here were focused on a conceptual design for a compact superconducting synchrocyclotron to demonstrate the possibility to further reduce its weight by almost another order of magnitude by eliminating all iron from the device. In the absence of magnetic iron poles, the magnetic field profile in the beam gap is achieved through a set of main superconducting split pair coils energized in series with a set of distributed field-shaping superconducting coils. External magnetic field shielding is achieved through a set of outer, superconducting ring coils, also connected in series with the other coils, to cancel the stray magnetic field. These shielding coils replace the heavy iron yoke which is the conventional method to return the magnetic flux. It is noted that the 10 Gauss surface is located at a radius of about 3.5 m comparable in both ironless and conventional devices, even in the absence of iron in the ironless device. An important result from eliminating all magnetic iron in the flux circuit is the resulting linear relationship between the operating current and the magnetic field intensity. In the case with iron, the saturation of the magnetic field forces operation at one value of magnetic field. This feature design then enables continuous beam energy variation without the use of an energy degrader, thus eliminating secondary radiation during the in-depth beam scanning, increasing the ion current delivered to the patient and improving the beam quality. The beam energy is determined by the magnetic field strength at the extraction radius, and changing the field enables selection of the final beam energy. The magnetic field can be adjusted while maintaining the needed radial field profile.