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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

The limits and challenges of error field correction for ITER

Significant progress has been made in interpreting the effects of non-axisymmetric 'error' fields on a plasma through ideal MHD stability and a dominant 'least stable' ideal mode through which the fields couple to the tearing resonant surface. However, in contrast to expectations from such theories, experiments have found limited success in correcting error fields, with single correction coil arrays giving benefits of between 0% and {approx}50% correction (in terms of improvement to a low density locked mode limit), dependent on the structure of the error and correcting fields. With additional coils up to {approx}70% is possible. It was unclear whether this represented an intrinsic stability or control limit, or higher order toroidal or poloidal harmonic effects. Thus, studies on the DIII-D tokamak explored correction of a proxy error field, using two differently structured coil arrays. This enabled the principles of error correction to be tested at high amplitudes and operational densities, with known pure n=1 fields. Results showed substantial residual effects from the corrected n=1 field, with improvements of only {approx}50% in the low density locked mode limit. This suggests that n=1 error fields must couple to more than one surface in the plasma, and this is conjectured to be through more than one ideal mode, thereby requiring precise correction. For ITER, updated predictions of field error have been obtained and compared with revised scalings for tearing mode thresholds, indicating 50% or better error field correction will be needed. This will likely require more than one well coupled correction coil array and sets a challenge for theory to model the behavior, in order to clarify the plasma response and braking mechanisms, and so the effectiveness of ITER's correction coils and the possible need for support from its edge localized mode control coils.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Tailoring tokamak error fields to control plasma instabilities and transport

A tokamak relies on the axisymmetric magnetic fields to confine fusion plasmas and aims to deliver sustainable and clean energy. However, misalignments arise inevitably in the tokamak construction, leading to small asymmetries in the magnetic field known as error fields (EFs). The EFs have been a major concern in the tokamak approaches because small EFs, even less than 0.1%, can drive a plasma disruption. Meanwhile, the EFs in the tokamak can be favorably used for controlling plasma instabilities, such as edge-localized modes (ELMs). Here we show an optimization that tailors the EFs to maintain an edge 3D response for ELM control with a minimized core 3D response to avoid plasma disruption and unnecessary confinement degradation. We design and demonstrate such an edge-localized 3D response in the KSTAR facility, benefiting from its unique flexibility to change many degrees of freedom in the 3D coil space for the various fusion plasma regimes. This favorable control of the tokamak EF represents a notable advance for designing intrinsically 3D tokamaks to optimize stability and confinement for next-step fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

The importance of matched poloidal spectra to error field correction in DIII-D

Optimal error field correction (EFC) is thought to be achieved when coupling to the least-stable “dominant” mode of the plasma is nulled at each toroidal mode number (n). The limit of this picture is tested in the DIII-D tokamak by applying superpositions of in- and ex-vessel coil set n = 1 fields calculated to be fully orthogonal to the n = 1 dominant mode. In co-rotating H-mode and low-density Ohmic scenarios, the plasma is found to be, respectively, 7× and 20× less sensitive to the orthogonal field as compared to the in-vessel coil set field. For the scenarios investigated, any geometry of EFC coil can thus recover a strong majority of the detrimental effect introduced by the n = 1 error field. Despite low sensitivity to the orthogonal field, its optimization in H-mode is shown to be consistent with minimizing the neoclassical toroidal viscosity torque and not the higher-order n = 1 mode coupling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Measurements and correction of the 1/1 error field in Wendelstein 7-X

In this paper, vacuum flux surface mapping is used to measure the error field in Wendelstein 7-X. Two methods for discriminating the perturbation from other co-resonant modes are applied: measurements of the helical axis shift in a magnetic configuration with the rotational transform barely above one on the axis; and measurements of the edge island structure with cancellation of intrinsic islands by divertor control coils. The normalized radial amplitude in Boozer coordinates is found to be about in the first case and about in the second one, with the discrepancy being attributable to different currents in the planar coils. The error field can be reliably compensated by using error field trim coils with currents about 100 A. Here, this compensation also helps to identify the error field of about . During plasma operation in divertor configurations, it is confirmed that the present field leads to asymmetries in the power distribution and that the compensation improves this symmetry.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Error field and magnetic diagnostic modeling for W7-X

The prediction, detection, and compensation of error fields for the W7-X device will play a key role in achieving a high beta (Β = 5%), steady state (30 minute pulse) operating regime utilizing the island divertor system [1]. Additionally, detection and control of the equilibrium magnetic structure in the scrape-off layer will be necessary in the long-pulse campaign as bootstrapcurrent evolution may result in poor edge magnetic structure [2]. An SVD analysis of the magnetic diagnostics set indicates an ability to measure the toroidal current and stored energy, while profile variations go undetected in the magnetic diagnostics. An additional set of magnetic diagnostics is proposed which improves the ability to constrain the equilibrium current and pressure profiles. However, even with the ability to accurately measure equilibrium parameters, the presence of error fields can modify both the plasma response and diverter magnetic field structures in unfavorable ways. Vacuum flux surface mapping experiments allow for direct measurement of these modifications to magnetic structure. The ability to conduct such an experiment is a unique feature of stellarators. The trim coils may then be used to forward model the effect of an applied n = 1 error field. This allows the determination of lower limits for the detection of error field amplitude and phase using flux surface mapping. *Research supported by the U.S. DOE under Contract No. DE-AC02-09CH11466 with Princeton University.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Error field correction strategies in preparation to MAST-U operation

In magnetic fusion devices, unwanted non-axisymmetric magnetic field perturbations, known as error fields (EF), can have detrimental effects on plasma stability and confinement. Such EFs may originate from several sources, such as axi-symmetric coil misalignments, coil feeds, 3D structures in the wall surrounding the plasma, presence of ferromagnetic materials near the plasma surface, blanket materials in future devices. To minimize their impact on plasma performance and on the available operational space, it is important to identify the EF sources and develop EF control strategies. MAST Upgrade (MAST-U) is a spherical tokamak which will operate in the near future after a series of enhancements from the previous MAST experiment. To deliver a machine with EF amplitude low enough to allow a high quality experimental programme, systematic analysis of the intrinsic EF sources has been carried out for poloidal eld (P) and divertor (D) coils. To deliver a machine with EF amplitude low enough to allow a high quality experimental programme, systematic analysis of the intrinsic EF sources has been carried out for poloidal field (P) and divertor (D) coils, whose magnetic eld measurements were available when writing this paper. Such measurements reveal that P and D coils are 3D deformed and thus are responsible for intrinsic EFs, with mainly n=1 and n=2 toroidal mode numbers. In preparation to MAST-U operation, both passive and active EF control strategies have been adopted for n=1 and n=2 EFs compensation. Passive EF control consisted of installing finely each D and P coil within MAST-U device so as to minimize the intrinsic n=1 EF amplitude. The optimal coil alignment has been determined based on magnetic eld measurements and the corresponding 3D electro-magnetic modelling, and envisaged coil shift and tilt of 3.2 mm and 0.7 mrad, respectively, in the case of P coil named P4. Conversely, active control will be use during MAST-U operation to reduce the n=2 EF which is associated mainly with P4 and P5, as well. Since these coils have been re-used from the MAST device, studies attempting n=2 EF control, based on MAST plasmas, have been modelled utilizing the MARS-F code to interpret experimental results in MAST and to give hints for future n=2 EF control studies in MAST-U. Here, a model-based control set for n=2 EF control has been identified which would allow for the minimization of rotation braking, of the resonant magnetic field at the q=2 and of the plasma displacement in MAST-U, simultaneously.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

The spectral basis of optimal error field correction on DIII-D

Here, experimental optimum error field correction (EFC) currents found in a wide breadth of dedicated experiments on DIII-D are shown to be consistent with the currents required to null the poloidal harmonics of the vacuum field which drive the kink mode near the plasma edge. This allows the identification of empirical metrics which predict optimal EFC currents with accuracy comparable to that of first- principles modeling which includes the ideal plasma response. While further metric refinements are desirable, this work suggests optimal EFC currents can be effectively fed-forward based purely on knowledge of the vacuum error field and basic equilibrium properties which are routinely calculated in real-time.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Nonlinear neoclassical two-fluid theory of response of tokamak plasma to resonant error-field

A nonlinear, neoclassical, two-fluid theory of the interaction of a single-helicity magnetic island chain with a resonant error-field in a quasi-cylindrical, low-β, tokamak plasma is presented. In particular, the analysis of Fitzpatrick [Phys. Plasmas 25, 042503 (2018)] is generalized to take explicit time dependence into account. Aside from the ability to more accurately treat time-varying problems, the main physical effect that is introduced into the theory by the incorporation of explicit time dependence is ion inertia. The formalism developed in the paper is used to analyze two time-varying problems. First, the interaction of a pre-existing magnetic island chain with a resonant error-field. Second, an error-field-maintained magnetic island chain. The latter problem is of direct relevance to experiments in which deliberately applied, multi-harmonic, resonant magnetic perturbations are used to suppress edge localized modes (ELMs) in tokamak plasmas. Indeed, the predictions of the theory are strikingly similar to data recently obtained from ELM suppression experiments in the DIII-D tokamak [R. Nazikian et al., Nucl. Fusion 58, 106010 (2018)].

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

An Investigation of Coupling of the Internal Kink Mode to Error Field Correction Coils in Tokamaks

The coupling of the internal kink to an external m/=1/1 pertubation if studied for profiles that are known to result in a saturated internal kink in the limit of a cylindrical tokamak. It is found from 3D equilibrium calculations that, for A 30 circular plasmas and A=3 elliptical shapes, this coupling of the boundary perturbation to the internal kink is strong; the amplitude of the m/n=1/1 structure at q=1 is large compared to the amplitude applied at the plasma boundary. It is proposed that this excitation, which could readily be applied with error field correction coils, be explored as a mechanism for controlling sawtooth amplitudes in high performance tokamak discharges. This saturated internal kink, resulting from small field errors in proposed as an explanation for the TEXTOR measurements of q0 and the distinction between sawtooth effects on the q-profile observed in TEXTOR and DIII-D.

q0↗
DOE OSTI2015

Decoupled recovery of energy and momentum with correction of n = 2 error fields

Experiments applying known n = 2 “proxy” error fields (EFs) find that the rotation braking introduced by the proxy EF cannot be completely alleviated through optimal n = 2 correction with poorly matched poloidal spectra. This imperfect performance recovery demonstrates the importance of correcting multiple components of the n = 2 field spectrum and is in contrast to previous results with n = 1 EFs despite similar execution. Measured optimal n = 2 proxy EF correction currents are consistent with those required to null dominant mode coupling to the resonant surfaces and minimize the neoclassical toroidal viscosity (NTV) torque, calculated using ideal MHD plasma response computation. Unlike rotation braking, density pumpout can be fully corrected despite poorly matched spectra, indicating density pumpout is driven only by a single component proportional to the resonant coupling. Through precise n = 2 spectral control density pumpout and rotation braking can thus be decoupled. Rotation braking with n = 2 fields is also found to be proportional to the level of concurrent toroidal rotation, consistent with NTV theory. Lastly, plasmas with modest countercurrent rotation are insensitive to the n = 2 field with neither rotation braking nor density pumpout observed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Simultaneous iterative learning control of mode entrainment and error field

Here, it is shown that static error fields (EFs) can severely limit the maximum rotation frequency achievable in mode entrainment by applied rotating fields. It is also shown that the rotation non-uniformities caused by an EF can be used to diagnose and correct said EF in real time. Simulations using typical DIII-D data show that this can be achieved within a small number of mode rotation periods by an iterative learning control algorithm. Further to rapidly correcting the EF, this gives access to high entrainment frequencies that would not be accessible otherwise, and paves the way to rotational stabilization and improved mode control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Effect of error field correction coils on W7-X limiter loads

In the first campaign Wendelstein 7-X was operated with five poloidal graphite limiters installed stellarator symmetrically. In an ideal situation the power losses would be equally distributed between the limiters. The limiter shape was designed to smoothly distribute the heat flux over two strike lines. Vertically the strike lines are not uniform because of different connection lengths. In this paper it is demonstrated both numerically and experimentally that the heat flux distribution can be significantly changed by non-resonant $n=1$ perturbation field of the order of $10^{-4}$ . Numerical studies are performed with field line tracing. In experiments perturbation fields are excited with five error field trim coils. The limiters are diagnosed with infrared cameras, neutral gas pressure gauges, thermocouples and spectroscopic diagnostics. Experimental results are qualitatively consistent with the simulations. With a suitable choice of the phase and amplitude of the perturbation a more symmetric plasma-limiter interaction can be potentially achieved. Finally, these results are also of interest for the later W7-X divertor operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Error field identification through torque balance on a saturated island in DIII-D

Measurement of the electromagnetic torque on a magnetic island could be an attractive method for error field identification in the early phase of ITER operation. Previous DIII-D experiments (Strait 2014 Nucl. Fusion 54 073004; Shiraki et al 2015 Plasma Phys. Control. Fusion 57 025016) have demonstrated the principle of this approach using a stationary or slowly rotating island, while recent developments in magnetic data analysis (Sweeney and Strait 2019 Phys. Plasmas 26 012509) allow the field of a rapidly rotating island to be readily distinguished from that of the wall currents induced by its rotation. In a recent experiment, a rotating n = 1 magnetic perturbation forced a saturated magnetic island to rotate, thus sampling all toroidal phases periodically in a single discharge. The phase and amplitude of the error field are inferred from analysis of the time-dependent torque balance on the island, including torques from the error field, the applied magnetic perturbation, and the wall currents induced by rotation of the applied perturbation and the island. Furthermore, results agree well with those from more conventional methods.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

JLEIC electron ring dynamic aperture with non-linear field errors

We present results of dynamic aperture study for the updated electron ring lattice of the Jefferson Lab Electron- Ion Collider (JLEIC). The lattice design features low emittance arcs with local compensation of sextupole nonlinear effects, and low emittance non-linear chromaticity correction sections. Dynamic aperture tracking simulations are performed to evaluate the effects of non-linear field errors, the sensitivity to betatron tune, and the impact of momentum error. Dynamic aperture is also evaluated with the measured PEP-II field errors. Preliminary tolerances to the non-linear field errors in the Final Focus quadrupoles are estimated.

Nosochkov, Yuri↗
DOE OSTI2022

Non-disruptive error field identification based on magnetic island healing

Here a technique to identify intrinsic error fields (EFs) in tokamaks with minimized risk of disruption is demonstrated on the DIII-D tokamak. The method extends the conventional driven magnetic island ‘compass scan’ approach by modifying asynchronous control waveforms to enable prompt healing of the island instability. Healing of the island is achieved by reducing the imposed non-axisymmetric coil current and raising the density (here via gas fueling). The method is also shown to support multiple island threshold measurements per pulse, thus reducing the number of dedicated pulses necessary to conduct an EF identification. Non-linear modeling with the TM1 code reproduces the experimental results and approximately recovers the critical density required for island healing. Island healing is explained in the non-linear modeling by an increase in the viscous coupling between the static island and the nearby flowing plasma, thus healing the island as it accelerates into the plasma frame. Due to both simplicity and risk minimization, this technique is suitable for plasma-based EF identification in the early commissioning stages of future disruption-averse tokamaks such as ITER and SPARC.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Error-field penetration in reversed magnetic shear configurations

Error-field penetration in reversed magnetic shear (RMS) configurations is numerically investigated by using a two-dimensional resistive magnetohydrodynamic model in slab geometry. To explore different dynamic processes in locked modes, three equilibrium states are adopted. Stable, marginal, and unstable current profiles for double tearing modes are designed by varying the current intensity between two resonant surfaces separated by a certain distance. Further, the dynamic characteristics of locked modes in the three RMS states are identified, and the relevant physics mechanisms are elucidated. The scaling behavior of critical perturbation value with initial plasma velocity is numerically obtained, which obeys previously established relevant analytical theory in the viscoresistive regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Error field detection and correction studies towards ITER operation

In magnetic fusion devices, error field (EF) sources, spurious magnetic field perturbations, need to be identified and corrected for safe and stable (disruption-free) tokamak operation. Within Work Package Tokamak Exploitation RT04, a series of studies have been carried out to test the portability of the novel non-disruptive method, designed and tested in DIII-D (Paz-Soldan et al 2022 Nucl. Fusion62 126007), and to perform an assessment of model-based EF control strategies towards their applicability in ITER. In this paper, the lessons learned, the physical mechanism behind the magnetic island healing, which relies on enhanced viscous torque that acts against the static electro-magnetic torque, and the main control achievements are reported, together with the first design of the asynchronous EF correction current/density controller for ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Error field predictability and consequences for ITER

Abstract ITER coil tolerances are re-evaluated using the modern understanding of coupling to least-stable plasma modes and an updated center-line-traced model of ITER’s coil windings. This reassessment finds the tolerances to be conservative through a statistical, linear study of n = 1 error fields (EFs) due to tilted, shifted misplacements and nominal windings of central solenoid and poloidal field coils within tolerance. We also show that a model-based correction scheme remains effective even when metrology quality is sub-optimal, and compare this to projected empirical correction schemes. We begin with an analysis of the necessity of error field correction (EFC) for daily operation in ITER using scalign laws for the EF penetration threshold. We then consider the predictability of EF dominant mode overlap across early planned ITER scenarios and, as measuring EFs in high power scenarios can pose risks to the device, the potential for extrapolation to the ITER Baseline Scenario (IBS). We find that carefully designing a scenario matching currents proportionally to those of the IBS is far more important than plasma shape or profiles in accurately measuring an optimal correction current set.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗