Engineering topics

Lazerson, Samuel A.

Publications and source records attributed to Lazerson, Samuel A..

13 records

DOE OSTI2024

Simulation of a scintillator-based fast ion loss detector for steady-state operation in Wendelstein 7-X (invited)

A quantitative theoretical framework has been created to model neutral beam injection and fast ion losses in the Wendelstein 7-X (W7-X) stellarator, including a novel method to develop synthetic diagnostics for fast ion loss detectors (FILDs) of many types, such as scintillating and Faraday Cup FILDs. This is the first time that this has been done in stellarator geometry with this level of fidelity, providing a way for fast ion losses to be predicted more precisely in future stellarator experiments and in W7-X. Simulations of the signal seen by a Faraday Cup FILD have been completed for multiple W7-X plasmas and show close agreement with the measured signals. This method is now applied to an actively water-cooled, scintillator-based FILD, which is currently in development to measure the fast ion loss distribution in W7-X in greater detail. The design makes use of a double slit to measure energy-and-pitch-angle-resolved losses of both co-going and counter-going fast ions. The diagnostic, which can be inserted to different radial positions, has been designed to withstand steady-state heat fluxes of up to 120 kW/m 2 along with additional transient heat loads of 100 kW/m 2 lasting for up to 20 s at a time. Simulations of W7-X standard magnetic configuration show up to 8 × 10 13 (s –1 cm –2 ) ion fluxes onto the sensor from each neutral beam source and no signal from the counter-going slit. Finally, these simulations will help inform experimental proposals for future W7-X campaigns after installation of this diagnostic.

3D printing↗

DOE OSTI2020

Optimization of quasi-axisymmetric stellarators with varied elongation

An optimization study of Quasi-Axisymmetric (QA) stellarators with varied elongation has been carried out using the optimization code STELLOPT. The starting point of our optimization is a previously obtained QA stellarator with 3 field periods and aspect ratio of 6. Here, a series of QA stellarators are obtained at zero plasma beta with varied elongation value ranging from 2.5 to 3.7. Good quasi-symmetry is kept when the elongation value is reduced from the original value of 3.7. The rotational transform profile and aspect ratio are kept fixed. The plasma volume is ether kept fixed or varied linearly with elongation. Furthermore, finite beta QA stellarators are considered. The corresponding bootstrap currents are calculated using the kinetic code SFINCS. A series of kink-stable QA stellarators are obtained via optimization with varied plasma beta up to 5% and self-consistent bootstrap current. This work demonstrates that good QA stellarators with finite beta and varied elongation exist that are stable to external kink modes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Investigation of the neoclassical ambipolar electric field in ion-root plasmas on W7-X

The role of the radial electric field in high-performance ion-root plasmas on Wendelstein 7-X (W7-X) is examined and compared with neoclassical predictions herein. The W7-X stellarator is the world’s first large-scale optimized stellarator. One of the important targets chosen for optimization during the W7-X design process was the reduction of core neoclassical heat transport. This optimization was targeted for reactor-relevant high-density plasmas with $T_e ≈ T_i$ in which the neoclassical ambipolar radial electric field is expected to be negative throughout the plasmas core. Measurements of the core radial electric field $(E_r)$ have confirmed that ion-root conditions (negative $(E_r)$ in the plasma core) have been achieved in W7-X with high-density plasmas and central ECRH. These measured $(E_r)$ profiles agree well with the neoclassical ambipolar $(E_r)$ predicted by the code SFINCS . This good agreement provides confidence in the validity of neoclassical calculations in high-density ion-root conditions, and enables initial studies on the role of neoclassical transport in the optimized high-density regime of W7-X. Profile measurements of electron temperature $(T_e)$, ion temperature $(T_i)$ and electron density $(n_e)$ along with approximations for the average value of $Z$ eff have been used as inputs to the sfincs code to calculate the ambipolar $E_r$ profile along with neoclassical ion and electron energy flux profiles $(Q$ NCi , $Q$ NCe $)$. Finally the total experimental energy input to the plasma from ECRH heating has been compared to the neoclassical energy fluxes to provide a first estimate for the fraction of transport that can be attributed to neoclassical processes in reactorrelevant high-density ion-root conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Tuning of the rotational transform in Wendelstein 7-X

The control of rotational transform in Wendelstein 7-X (W7-X) is key to the island divertor operation and safety of plasma facing components. The island divertor concept in W7-X relies on an edge flux surface with rotational transform of $t$ = 1 resonating with an intrinsic n/m = 5/5 resonance to form a five lobed island chain. This island chain intersects with divertor plates to give rise to the island divertor. Changes in the relative position of the rational surface and the divertor plates can result in changes in divertor performance, thus the control of the rotational transform is essential to operation of the W7-X device. During the first divertor campaign electromagnetic loads resulted in elastic deformations of the shaped modular stellarator coils. Such deformations made these coils more planar, reducing the vacuum rotational transform, subsequently shifting the $t$ = 1 resonance outward. Unintended plasma wall interactions provided the first clear evidence of this effect during plasma operation. Flux surface measurements were utilized to estimate the correct level of current in the planar coils for correction of $t$, and found to be around ~-100 A . Scans of the planar coil current for iota correction were performed during plasma operation. These measurements suggest planar coil currents between -250 and ~ 0 A would place the strike lines at the designed distance from the pumping gaps. Divertor Langmuir and upstream probe measurements confirm these estimates along with measurements of divertor neutral gas pressure.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Development of a Faraday cup fast ion loss detector for keV beam ions

The development and testing of a Faraday cup fast-ion loss detector capable of measuring sub 100 keV particles is documented. Such measurement capabilities play an important role in the assessment of particle confinement of nuclear fusion experiments. The detector is manufactured using thin-film deposition techniques, building upon previous work using discrete foils. This new manufacturing method allows the form factor of the sensor to become that of essentially a microchip. Analysis of the diagnostic response is performed using Monte-Carlo particle simulations. These simulations show peaks in the detector response at 40 and 70 keV. The sensor is then tested in a tunable linear accelerator capable of accelerating protons from 20 to 120 keV. Finally, the detector response was found to be well matched to simulations. Improvements to the design to facilitate robustness are discussed.

47 OTHER INSTRUMENTATION↗

DOE OSTI2019

Ion temperature gradient turbulence modification in quasi-axisymmetry

The large flexibility of the proposed QUASAR facility is leveraged in order to explore the effect of magnetic shear on adiabatic Ion Temperature Gradient (ITG) turbulence. The QUASAR facility is a reimagining of the National Compact Stellarator Experiment utilizing and expanding upon the already constructed coil set. Recent work using fixed boundary optimization of the LI383 equilibrium (upon which QUASAR is based) has suggested possible improvements to ITG turbulence. In this work, a different approach is taken, wherein a series of self-consistent free boundary VMEC equilibria are developed for QUASAR. These equilibria assume temperature and density profiles consistent with 2% beta and ohmic current drive. In each configuration, the toroidal field coils are energized to different values and the STELLOPT code is used to vary the modular coil current and net toroidal current. The edge value of rotational transform is targeted in the optimization, producing a magnetic shear scan. All these configurations share similar neoclassical transport levels, while nonlinear GENE flux tube simulations show up to a factor of four change in adiabatic ITG turbulence at various radii. Comparisons of proxy functions and linear flux tube runs are also made. This work highlights the capability of the QUASAR experiment as a tool to explore transport in 3D magnetic fields and the possibility of the further improvements to stellarators through optimization.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2018

Error fields in the Wendelstein 7-X stellarator

The role of error fields in stellarators is explored in this work through simulations of, and experiments performed on, the Wendelstein 7-X (W7-X) stellarator. Previous experimental results, based on limiter plasmas in W7-X, suggest that error fields are small in the experiment based on limiter plasmas. In this work, we document results from the first divertor campaign on W7-X with a focus on error fields resonant with the edge island divertor structure. Measurements obtained using divertor thermocouples corroborate earlier limiter results that the error fields are small and correctable, requiring only 5% of the total rated current of the trim coil system. The effect of these error fields on magnetic geometry and topology are investigated using the flux surface measurement technique. The resulting divertor thermal load asymmetry is modeled via field line diffusion and compared to experimental data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2018

Hessian matrix approach for determining error field sensitivity to coil deviations.

The presence of error fields has been shown to degrade plasma confinement and drive instabilities. Error fields can arise from many sources, but are predominantly attributed to deviations in the coil geometry. In this paper, we introduce a Hessian matrix approach for determining error field sensitivity to coil deviations. A primary cost function used for designing stellarator coils, the surface integral of normalized normal field errors, was adopted to evaluate the deviation of the generated magnetic field from the desired magnetic field. The FOCUS code [Zhu et al., Nucl. Fusion 58(1):016008 (2018)] is utilized to provide fast and accurate calculations of the Hessian. The sensitivities of error fields to coil displacements are then determined by the eigenvalues of the Hessian matrix. A proof-of-principle example is given on a CNT-like configuration. We anticipate that this new method could provide information to avoid dominant coil misalignments and simplify coil designs for stellarators.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2017

Error field measurement, correction and heat flux balancing on Wendelstein 7-X

The measurement and correction of error fields in Wendelstein 7-X (W7-X) is critical to long pulse high beta operation, as small error fields may cause overloading of divertor plates in some configurations. Accordingly, as part of a broad collaborative effort, the detection and correction of error fields on the W7-X experiment has been performed using the trim coil system in conjunction with the flux surface mapping diagnostic and high resolution infrared camera. In the early commissioning phase of the experiment, the trim coils were used to open an n/m = 1/2 island chain in a specially designed magnetic configuration. The flux surfacing mapping diagnostic was then able to directly image the magnetic topology of the experiment, allowing the inference of a small similar to 4 cm intrinsic island chain. The suspected main sources of the error field, slight misalignment and deformations of the superconducting coils, are then confirmed through experimental modeling using the detailed measurements of the coil positions. Observations of the limiters temperatures in module 5 shows a clear dependence of the limiter heat flux pattern as the perturbing fields are rotated. Plasma experiments without applied correcting fields show a significant asymmetry in neutral pressure (centered in module 4) and light emission (visible, H-alpha, CII, and CIII). Such pressure asymmetry is associated with plasma-wall (limiter) interaction asymmetries between the modules. Application of trim coil fields with n = 1 waveform correct the imbalance. Confirmation of the error fields allows the assessment of magnetic fields which resonate with the n/m = 5/5 island chain.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

First measurements of error fields on W7-X using flux surface mapping

Error fields have been detected and quantified using the flux surface mapping diagnostic system on Wendelstein 7-X (W7-X). A low-field '${\rlap{-}\ \iota} =1/2$ ' magnetic configuration (${\rlap{-}\ \iota} =\iota /2\pi $ ), sensitive to error fields, was developed in order to detect their presence using the flux surface mapping diagnostic. In this configuration, a vacuum flux surface with rotational transform of n/m = 1/2 is created at the mid-radius of the vacuum flux surfaces. If no error fields are present a vanishingly small n/m = 5/10 island chain should be present. Modeling indicates that if an n = 1 perturbing field is applied by the trim coils, a large n/m = 1/2 island chain will be opened. This island chain is used to create a perturbation large enough to be imaged by the diagnostic. Phase and amplitude scans of the applied field allow the measurement of a small $\sim 0.04$ m intrinsic island chain with a ${{130}^{\circ}}$ phase relative to the first module of the W7-X experiment. Lastly, these error fields are determined to be small and easily correctable by the trim coil system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Verification of the ideal magnetohydrodynamic response at rational surfaces in the VMEC code

The VMEC nonlinear ideal MHD equilibrium code [S. P. Hirshman and J. C. Whitson, Phys. Fluids 26, 3553 (1983)] is compared against analytic linear ideal MHD theory in a screw-pinch-like configuration. The focus of such analysis is to verify the ideal MHD response at magnetic surfaces which possess magnetic transform (ι) which is resonant with spectral values of the perturbed boundary harmonics. A large aspect ratio circular cross section zero-beta equilibrium is considered. This equilibrium possess a rational surface with safety factor q = 2 at a normalized flux value of 0.5. A small resonant boundary perturbation is introduced, exciting a response at the resonant rational surface. The code is found to capture the plasma response as predicted by a newly developed analytic theory that ensures the existence of nested flux surfaces by allowing for a jump in rotational transform (ι=1/q). The VMEC code satisfactorily reproduces these theoretical results without the necessity of an explicit transform discontinuity (Δι) at the rational surface. It is found that the response across the rational surfaces depends upon both radial grid resolution and local shear (dι/dΦ, where ι is the rotational transform and Φ the enclosed toroidal flux). Calculations of an implicit Δι suggest that it does not arise due to numerical artifacts (attributed to radial finite differences in VMEC) or existence conditions for flux surfaces as predicted by linear theory (minimum values of Δι). Scans of the rotational transform profile indicate that for experimentally relevant levels of transform shear the response becomes increasing localised. Furthermore, careful examination of a large experimental tokamak equilibrium, with applied resonant fields, indicates that this shielding response is present, suggesting the phenomena is not limited to this verification exercise.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2014

BEAMS3D Neutral Beam Injection Model

With the advent of applied 3D fields in Tokamaks and modern high performance stellarators, a need has arisen to address non-axisymmetric effects on neutral beam heating and fueling. We report on the development of a fully 3D neutral beam injection (NBI) model, BEAMS3D, which addresses this need by coupling 3D equilibria to a guiding center code capable of modeling neutral and charged particle trajectories across the separatrix and into the plasma core. Ionization, neutralization, charge-exchange, viscous velocity reduction, and pitch angle scattering are modeled with the ADAS atomic physics database. Benchmark calculations are presented to validate the collisionless particle orbits, neutral beam injection model, frictional drag, and pitch angle scattering effects. A calculation of neutral beam heating in the NCSX device is performed, highlighting the capability of the code to handle 3D magnetic fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2013

STELLOPT Modeling of the 3D Diagnostic Response in ITER

The ITER three dimensional diagnostic response to an n=3 resonant magnetic perturbation is modeled using the STELLOPT code. The in-vessel coils apply a resonant magnetic perturbation (RMP) fi eld which generates a 4 cm edge displacement from axisymmetry as modeled by the VMEC 3D equilibrium code. Forward modeling of flux loop and magnetic probe response with the DIAGNO code indicates up to 20 % changes in measured plasma signals. Simulated LIDAR measurements of electron temperature indicate 2 cm shifts on the low field side of the plasma. This suggests that the ITER diagnostic will be able to diagnose the 3D structure of the equilibria.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗