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At least 487 records · Page 27

Localizing resonant magnetic perturbations for edge localized mode control in KSTAR

An external 3D magnetic perturbation typically drives a resonant response at the rational surfaces from the core to the edge of tokamak plasmas, due to strong mode coupling and amplification. In this paper, we present a method to isolate the edge from core resonant fields using the ideal perturbed equilibrium code and to design an edge-localized resonant magnetic perturbation (RMP) for effective edge localized mode (ELM) control. A robust feature of the edge-localized RMP is the curtailed response to the field at the low-field-side (LFS) midplane, as opposed to typical RMPs which strongly resonate with the LFS fields. This emphasizes the importance of off-midplane coils to improve ELM control without provoking a large core response that could lead to devastating instabilities. The conceptual design of new ELM control coils based on the edge-localized RMP in KSTAR shows how this new insight can be utilized to enhance the efficiency of our ELM suppression capabilities. Simple window-pane coils matching the edge-localized resonant mode structure substantially expand in the ELM suppression window beyond the existing coil. Further optimization using the flexible optimized coils using space-curves code leads to additional enhancement in the edge-localized control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Performance of a MQXF Nb 3 Sn Quadrupole Magnet Under Different Stress Level

In a dipole or in a quadrupole accelerator magnet, the displacement of the coil turns induced by the electromagnetic forces can cause quenches limiting the magnet performance. For this reason, an azimuthal preload is applied to avoid azimuthal movements of the coil up to the required operational current. However, several tests showed that accelerator magnets can operate with a partial preload, i.e. that coil unloading during the ramp does not prevent reaching higher currents. This issue is particularly relevant for Nb 3 Sn magnets, where the loads applied to the Nb3Sn filaments can reach the degradation limits of critical current. In order to investigate the impact of coil preload on the quench performance, the MQXFS6 short model quadrupole for the High Luminosity Upgrade was tested under an azimuthal preload at 80% of the short sample current, reaching 93% of short sample current at 1.9 K. The preload was then released to 60%, still showing ability to operate in the range of 80-85% of short sample current as required by HL-LHC project. With this lower preload, the ability of going above 90% of short sample was lost, and a significant training appeared above 85%. When the preload was restored to the original 80% value, the magnet reached with few quenches 95% of short sample (13.4 T peak field). Magnetic measurements confirm the larger movement of the coil in the case with lower preload, and agree with finite element simulations.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Field Measurement Results of the 15 T Nb 3 Sn Dipole Demonstrator MDPCT1b

A 15 T Nb 3 Sn dipole with cos-theta-type coils was developed to demonstrate a possible magnet design for a post-LHC proton-proton Collider. The magnet has 4-layer coils with 60-mm aperture, and has graded current density between the inner and outer layers to maximize performance. The coils are constrained by vertically-split, thick iron laminations, connected by aluminum Iclamps, and a thick stainless-steel skin. The magnet was previously tested at the Vertical Magnet Test Facility (VMTF) at Fermilab with reduced pre-load, and achieved record 14.1 T fields at maximum current. Now the magnet has been reassembled with full pre-load and has been cryogenically retested at VMTF. We report here on the magnetic measurements results of these tests, including field strength, geometrical harmonics, coil magnetization, and iron saturation characteristics of the demonstrator.

43 PARTICLE ACCELERATORS↗

First ITER CS module test results

The ITER Central Solenoid (CS) is under fabrication by the US ITER organization and its subcontractors. US ITER will supply seven modules to ITER IO, six of which will be assembled in a stack that forms the ITER Central Solenoid, with one as a spare. The first module fabrication has been completed by General Atomics (GA) at their facility and has begun testing including high voltage testing, Paschen testing in the vacuum and then testing at 4.5 K and up to 40 kA in order to demonstrate compliance of the coil to ITER requirements. In the paper we present the Test Plan and results of the CS Module performance tests, especially at 40 kA current. Herein, AC losses, joint resistances and hydraulic characteristics of the coil are all measured. Displacements of the coil height and hoop strain of the CS Module are also measured to verify structural and mechanical characteristics of the coil along with cooldown shrinkage of the coil. This information is used for verification of the stack behavior of CS in ITER operation. The test results and preliminary analyses results are presented, compared to expectations, and discussed. All measured parameters suggest that the CS module will perform well in ITER machine.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Laser-driven quasi-static B-fields for magnetized high-energy-density experiments

We present measurements of magnetic fields generated in laser-driven coil targets irradiated by laser pulses of nanosecond duration, 1.053 μm wavelength, 500 J energy, and ∼1015 W/cm2 intensity, at the LULI2000 facility. Using two perpendicular probing axes, proton deflectometry is used to characterize the coil current and static charge at different times. Results reveal various deflection features that can be unambiguously linked to a looping quasi-steady current of well-understood polarity or to a static charging of the coil surface. Measured currents are broadly consistent with predictions from a laser-driven diode-current source and lumped circuit model, supporting the quasi-steady assessment of the discharges. Peak magnetic fields of ∼50 T at the center of 500-μm-diameter coils, obtained at the moderate laser intensity, open up the use of such laser-driven coil targets at facilities worldwide to study numerous phenomena in magnetized high-energy-density plasmas, and its potential applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Current potential patches

A novel form of the current potential, a mathematical tool for the design of stellarators and stellarator coils, is developed. Specifically, these are current potentials with a finite-element basis, called current potential patches. Current potential patches leverage the relationship between distributions of magnetic dipoles and current potentials to explore limits of the access properties of stellarator coil sets. An example calculation is shown using the Helically Symmetric Experiment (HSX) equilibrium, demonstrating the method's use in coil design and understanding the limits of the access properties of coil sets. Current potential patches have additional desirable properties such as of promoting sparse current sheet solutions and identifying crucial locations of shaping current placement. A result is found for the HSX equilibrium that shaping currents covering only 22% of the winding surface are sufficient to produce the equilibrium to a good accuracy, provided a toroidal field is generated by an exterior coil set.

Elder, Todd↗

Modeling plasma toroidal flow profile control via NTV torque with n = 2 3D fields in MAST-U

Toroidal modeling utilizing the MARS-Q code shows that large (compared to electromagnetic torque) neoclassical toroidal viscous (NTV) torque can be achieved in the plasma core region in reference MAST(-U) L-mode plasmas, by applying the n=2 (n is the toroidal mode number) 3-D fields generated by the magnetic coils used for controlling the edge localized mode and/or by the error field correction coils. Large NTV torque, occurring at relatively slow plasma flow, in combination with strong variation of the torque amplitude versus the coil phasing of the 3-D coils, offers a tool to control the plasma toroidal rotation profile in spherical tokamaks. To effectively employ the toroidal NTV torque, the potentially fast initial flow needs to first be damped by 3-D fields. In conclusion, MARS-Q quasi-linear initial value simulations demonstrate that this is achievable in MAST-U, within the designed 3-D coil current capability for the device.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Influence of elongation and triangularity on plasma response to resonant magnetic perturbations

Abstract Systematic toroidal modeling of the plasma response to the n = 1–4 ( n is the toroidal mode number) resonant magnetic perturbation (RMP) field is carried out in order to understand the plasma-shaping effect on controlling the type-I edge-localized modes (ELMs) in tokamak experiments. Considered are large variations of the plasma elongation and triangularity at a fixed edge safety factor q a , for limiter plasmas with both single-null (SN) and double-null (DN) divertor-like boundary shapes. Numerical results assuming conformal 3D RMP coils show that (i) the optimum coil phasing between the upper and lower rows for ELM control becomes increasingly sensitive to the plasma elongation with higher- n toroidal spectra, (ii) the optimum coil phasing is however essentially independent of the plasma triangularity for all n = 1–4 RMP fields, (iii) with the same coil current and the optimum coil phasing, high elongation generally favors ELM control but it may be more challenging for plasmas with intermediate elongation and with lower- n ( n = 1–2) RMPs, and (iv) higher triangularity is generally always better for ELM control with all n = 1–4 fields for both DN and SN divertor-like plasma boundary shapes.

Physics↗

Non-linear MHD modelling of edge localized modes suppression by resonant magnetic perturbations in ITER

Edge localized modes (ELMs) suppression by resonant magnetic perturbations (RMPs) was studied with the non-linear magneto-hydro-dynamic (MHD) code JOREK for the ITER H-mode scenarios at 15 MA, 12.5 MA, 10 MA/5.3 T. In this study, the main aim was to demonstrate that ELMs can be suppressed by RMPs while the divertor 3D footprints of heat and particle fluxes remain within divertor material limits. The unstable peeling–ballooning modes responsible for ELMs without RMPs were modelled first for each scenario using numerically accessible parameters for ITER. Then the stabilization of ELMs by RMPs was modelled with the same parameters. RMP spectra, optimized by the linear MHD MARS-F code, with main toroidal harmonics N = 2, N = 3, N = 4 have been used as boundary conditions of the computational domain of JOREK, including realistic RMP coils, main plasma, scrape off layer (SOL) divertor and realistic first wall. The model includes all relevant plasma flows: toroidal rotation, two fluid diamagnetic effects and neoclassical poloidal friction. With RMPs, the main toroidal harmonic and the non-linearly coupled harmonics remain dominant at the plasma edge, producing saturated modes and a continuous MHD turbulent transport thereby avoiding ELM crashes in all scenarios considered here. The threshold for ELM suppression was found at a maximum RMP coils current of 45 kAt–60 kAt compared to the coils maximum capability of 90 kAt. In the high beta poloidal steady-state 10 MA/5.3 T scenario, a rotating QH-mode without ELMs was observed even without RMPs. In this scenario with RMPs N = 3, N = 4 at 20 kAt maximum current in RMP coils, similar QH-mode behaviour was observed however with dominant edge harmonic corresponding to the main toroidal number of RMPs. The present MHD modelling was limited in time by few tens of ms after RMPs were switched on until the magnetic energy of the modes saturates. As a consequence the thermal energy was still evolving on this time scale, far from the ITER confinement time scale and hence only the form of 3D footprints on the divertor targets can be indicated within this set-up. Also note, that the divertor physics was missing in this model, so realistic values of fluxes are out of reach in this modelling. However the stationary 3D divertor and particle fluxes could be simply extrapolated from these results to the stationary situation considering that a large power fraction should be radiated in the core and SOL and only about 50 MW power is going to the divertor, which is an arbitrary, but reasonable number used here. The 3D footprints with RMPs show the characteristic splitting with the main RMP toroidal symmetry. The maximum radial extension of the footprints typically was ~20 cm in inner divertor and ~40 cm in outer divertor with stationary heat fluxes decreasing further out from the initial strike point from ~5 MW m ₋2 to ~1 MW m ₋2 assuming a total power in the divertor and walls is 50 MW. The heat fluxes remain within the divertor target and baffle areas, however with rather small margin in the outer divertor which could be an issue for the first wall especially in transient regimes when part of the plasma thermal energy is released due to switching on the RMP coils. This fact should be considered when RMPs are applied with a more favorable application before or soon after the L–H transition, although optimization is required to avoid increasing the L–H power threshold with RMPs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Quasi-linear toroidal simulations of resonant magnetic perturbations in eight ITER H-mode scenarios

Abstract Both linear and quasi-linear aspects of the plasma response to the resonant magnetic perturbation (RMP) field are numerically investigated for various H-mode scenarios in ITER, covering the pre-fusion power operation and the fusion power operation phases. Linear response computations for eight ITER scenarios, with varying plasma current and toroidal magnetic field, reveal that the best coil current phasing for controlling the type-I edge localized modes (ELMs) scales roughly linearly with the edge safety factor. The coil phasing is defined as the relative toroidal phase of the coil currents between different rows, for a given toroidal harmonic. Quasi-linear initial value simulation, which is the focus of the present study, shows that application of the n = 3 ( n is the toroidal mode number) RMP field has a minimum side effect on the plasma core momentum confinement but potentially a large effect on the global particle transport. Generally, the RMP field with the best (worst) coil phasing for ELM control produces the strongest (weakest) effect on the plasma edge flow and the overall density. This robustly holds for all eight ITER scenarios. Consequently, in order to minimize the RMP induced side effects while achieving ELM control (suppression) in ITER, a compromise is necessary in choosing the coil current configuration.

Physics↗

Multi-machine validation of plasma initiation modelling and prospects for future devices: Predicting plasma initiation using only hardware design and control room input data

This paper reports on the generic prediction capability of full electromagnetic plasma initiation modelling with DYON, which was carried out for the first time in fusion research by the joint modelling of the International Tokamak Physics Activity—Integrating Operation Scenario group. The following devices were included in the experiment database: VEST (spherical torus, copper coils, Stainless steel wall, R/a = 0.3 m/0.2 m, V v = 3.7 m 3 ), MAST-U (spherical torus, copper coils, C wall, R/a = 0.7 m/0.5 m, V v = 55 m 3 ), EAST (conventional tokamak, superconducting coils, metallic wall, R/a = 1.85 m/0.5 m, V v = 38 m 3 ), DIII-D (conventional tokamak, copper coils, C wall, R/a = 1.67 m/0.65 m, V v = 35 m 3 ), and KSTAR (conventional tokamak, superconducting coils, C wall, R/a = 1.8 m/0.5 m, V v = 55 m 3 ). Despite the different hardware features of the devices, the required operating spaces of the loop voltage induction and prefill gas pressure for inductive plasma initiation in each device were successfully reproduced by the predictive simulations with DYON using only the individual hardware design and the control room input data for each discharge. This successful validation across multiple machines demonstrates that the full electromagnetic DYON modelling can capture the essential physics of inductive plasma initiation. The simulation settings commonly employed for all modelling and the modifications necessary to account for the discrepancies between individual devices are reported. Predictions for ITER based on the multi-machine validation indicate that a wide range of prefill gas pressures exists for the Townsend breakdown and the plasma burn-through (0.01–1.5 mPa).

DYON↗

Permanent magnets for ELM suppression in tokamaks: feasibility and operational compatibility

Permanent magnets can provide static magnetic fields without power supplies or feed lines, offering a simpler alternative to conventional electromagnets in tokamaks. This work examines permanent magnet arrays (PMAs) for edge localized mode (ELM) control in double-null (DN) configurations, where traditional RMP coils exhibit limited effectiveness. Linear plasma response calculations using IPEC assess high-field-side (HFS) permanent magnet placement, benchmarking performance against DIII-D low-field-side (LFS) internal coils (I-coils) while evaluating engineering constraints. Modeling results demonstrate that HFS permanent magnet configurations generate HFS plasma response amplitudes more than 5 times larger than conventional I-coil systems. While other resonant response metrics show comparable or reduced response relative to I-coils, the combination of conventional RMP systems and permanent magnets is expected to provide enhanced performance. Operational impact assessments of persistent magnetic fields, including sideband field effects and startup/ramp-up compatibility, reveal acceptable performance within established operational boundaries. This analysis establishes PMAs as a technically viable approach for DN ELM control with reductions in system complexity, motivating further experimental validation on existing tokamak facilities.

3D magnetic perturbation↗

HTS Quadrupole Magnet for the Persistent Current Mode Operation

In this work, a new high-temperature superconducting (HTS) quadrupole magnet with circular coils was designed and built at Fermilab. Several HTS coils were also investigated at the temperature of liquid nitrogen. The main goal of this activity was to investigate the coils and magnet operation in a persistent current mode to reduce accelerator magnets’ capital and operational expenses. To accomplish this, HTS short-circuit coils were used. In the paper, HTS coils and magnet design, fabrication, and tests were discussed. The magnetic field in the magnet aperture and the field decay when the magnet operated in the “frozen flux” mode were measured. The test and simulation results were compared, confirming the advantages of the proposed approach.

43 PARTICLE ACCELERATORS↗

Final Design and Analysis of the Superconducting Magnets for the Material Plasma Exposure Experiment

The Material Plasma Exposure eXperiment (MPEX), which has completed its final design, is a new linear plasma device to advance the understanding of plasma-material interactions through the generation and delivery of plasmas as they are expected in future fusion reactor divertors. MPEX will be a steady-state device to study high-fluence exposures of plasma-facing materials and components. The requirements for the magnetic field at the heating stages and the target make the application of superconducting coils necessary. The final designs for the superconducting magnets have been developed using three cryostat designs with warm bore diameters of 65 cm and 156 cm. The large bores are required to for other systems such as vacuum, water cooling, and RF power. There are 19 superconducting coils in MPEX that are contained in six cryostats. Although design, fabrication, and testing for the magnets as stand-alone units are straightforward, challenges will arise during the integration of the system. Various field profiles will be used during operation. The magnetic field where the electron cyclotron heating occurs needs to operate at both 1.25 and 2.50 T. Analyses that have been performed on the final design include designing the coils and determining operating currents to meet field requirements, forces between cryostats, the effects of coil and/or cryostat movement, and quench analysis. To ensure that the magnetic field requirements are met, a plan for monitoring the magnetic field at specified locations has been developed which includes the effect of coil and/or cryostat movement.

Burkhardt, Earle↗

Testing of the ITER Central Solenoid Modules

The ITER Central Solenoid is under fabrication by the U.S. ITER organization and its subcontractors. U.S. ITER will supply seven modules to ITER IO, six of which will be assembled in a stack that forms the ITER Central Solenoid. The first modules that were built by GA at their facility, went into high voltage testing, including Paschen testing in the vacuum, and then they were tested at 4.5 K and up to 40 kA to demonstrate compliance of the coil with the ITER requirements. In this article, we present the Test Plan and results of the central solenoid (CS) module’s performance, especially at the full current. We measured critical temperatures in several pancakes, we measured ac losses, joint resistance, and hydraulic characteristics of the coils. We also measured displacements of the coil height and hoop strain of the CS module (CSM) to verify the structural mechanical characteristics of the coil along with the cooldown shrinkage of the coil. We studied the performance of the cowound quench detectors and confirmed their effectiveness in the suppression of inductive noise. This information is necessary for verification of the stack behavior of CS in ITER operation. The test results and preliminary analyses are presented, compared to expectations, and discussed.

Martovetsky, Nicolai↗

An Integrated Electric Vehicle Drive Motor and Wireless Charger

This work proposes an in-wheel electric vehicle-motor in which the motor windings also act as receiver coils in a wireless charging system. Vehicle wireless charging systems normally include a receiver coil, compensation network, rectifier and battery charger. The proposed concept would eliminate the need for a separate receiver coil. Studies based on two-dimensional finite element analysis are carried out to verify the feasibility of the concept. An external rotor Halbach array rotor with single layer fractional slot concentrated wingdings is found to be a suitable motor topology which would enable the use of one or more motor coils as a receiver coil. The paper also discusses the design modifications and considerations required to enable this.

Rallabandi, Vandana↗

Plasma Guns for Magnetized Fuel Targets for PJMIF

In plasma jet driven magneto-inertial fusion (PJMIF) an array of discrete supersonic plasma jets is used to form a spherically imploding plasma liner, which then compresses a magnetized plasma target to fusion conditions. With funding from ARPA-E’s ALPHA program and from Strong Atomics LLC, HyperJet Fusion Corp and HyperV Technologies Corp. previously developed the plasma guns required for an experimental demonstration of the plasma liner formation part of the concept. A 36-gun demonstration of an imploding spherical plasma liner is currently underway on the PLX facility at Los Alamos National Laboratory. This present project addresses the next step required for a complete PJMIF concept, i.e. developing the magnetized plasma target. We proposed to form the target by stagnating a number of magnetized plasma jets in the center of the target chamber. This is accomplished by adding a bias field coil to the plasma liner gun to form a magnetized plasma jet. This experimental development took place at HyperJet in a geometry replicating the bias field environment that will be seen on a PLX port so the results are directly transferrable to the PLX experiment. The objective of this effort was the technical development and characterization of a new magnetized plasma jet using a high-performance, high momentum flux, contoured-gap coaxial plasma gun appropriate for use on the next stage of the PLX experiment. Electromagnetic modeling of the coil indicated the coil was best placed around the aluminum tubes of the gun transmission line, rather than around the chamber port as originally proposed. This maximized field strength in the breech and yielded much better flux linkage between gun electrodes. A 30-turn coil was ultimately implemented, allowing long pulses that could diffuse through the metal walls on the timescale of interest. \machtwo modeling predicted that less capacitance in the main PFN could potentially improve plasma jet velocities due to better matching of the current to the smaller plasma mass and the existing electrode contour designed to suppress blowby. This proved true, as testing showed markedly improved performance when the original 600uF bank was reduced to 400uF. A number of diagnostics were built and/or upgraded in order to characterize the plasma jets, inluding laser interferomtery for density, photodiodes for velocity, Bdot probes for magnetic field measurements, a Triple probe for temperature measurements, and spectroscopy for impurity content. A plasma gun with the 30-turn magnet coil installed, a 70% reduction in gas valve plenum volume, and a 33% reduction in main PFN capacitance produced a dense, high velocity, well magnetized plasma jet. We met or exceeded virtually all of the plasma jet parameter goals. Peak velocity of 135 km/s exceeded the 100 km/s goal by 35%, while the peak density of over 1.0x10 15 cm -3 was 3.3 times the goal of >3.0x10 14 cm -3 . Shot-to-shot repeatability is excellent, with a jitter of less than 300ns observed on the arrival fronts of the photodiode signals from one shot to the next. Average plasma jet lengths of 33cm (at 120km/s) were slightly longer than the 20cm goal, but jets as short as 11.8cm were observed at 135km/s. Mass is much higher than the targeted goal, averaging 106ug per shot compared to 20ug. The magnetic field is still a bit lower than desired, with a maximum to date of ~811G, compared to the goal of 1000G. Average values, though, were typically in the 300-450G range, when measured further downstream after some expected in flight decay. Temperature measurements are still a work in progress. Increasing the B field and completing temperature measurements will be continued on into the ongoing BETHE project.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Experimental Investigation of a Constant-Velocity Traveling Magnetic Wave Plasma Engine

The previously reported traveling magnetic wave plasma engine has been redesigned to reduce the high heat loss to the tube walls downstream of the last magnetic-field coil. The present configuration uses a 3-inch-diameter pyrex tube flared out to a 6-inch diameter immediately downstream of the last magnetic-field coil. This configuration has been studied to determine the effects of flared-tube geometry, molecular weight of the propellant, engine length, and use of a ferrite core on the engine performance. Argon and xenon gases were used as propellants. Two engine lengths were tested. One engine was nominally 1-magnetic-wavelength long (4 coils) and the other nominally 2 1/2-magnetic wavelengths long (10 coils). The magnetic wave speed for both of these lengths corresponds to a specific impulse of 4750 seconds. The maximum kinetic efficiency of the 4-coil engine was 10 percent at a specific impulse of 3200 seconds using argon gas and 22.5 percent at 4200 seconds using xenon gas as the propellant.

TRAVELING WAVE↗