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Meyer, W. H.

Publications and source records attributed to Meyer, W. H..

14 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Validation of EDGDE2D-EIRENE predicted 2D distributions of electron temperature and density against divertor Thomson scattering measurements in the low-field side divertor leg in DIII-D

Comparisons of profiles of the electron temperature (T e ), density (n e ) and pressure (p e ) measured with Divertor Thomson Scattering in DIII-D low-confinement mode discharges to predictions from the edge fluid code EDGE2D-EIRENE show that the models implemented in EDGE2D-EIRENE predict the measurements within their collective uncertainties if the T e at the separatrix (T e,sep ) is 10 eV, or higher. The simulations do not predict, however, the peaked T e and ne profiles measured adjacent to the target plate when T e,sep , is below 10 eV, i.e., for the plasma downstream from the region of ionization of deuterium atoms. Inclusion of cross-field drifts and a five-fold reduction of radial transport cannot reconcile the discrepancy between the measurements and predictions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Implementation and assessment of an extended hydrogenic molecular model in UEDGE

Abstract Two sets of pure deuterium plasmas are simulated using UEDGE, one including atoms and molecules and the other including atoms only as separate fluid species. Simulation results in one and two dimensions are reported to assess the role of molecules in tokamak plasmas. It is shown that thermal coupling of the molecules to the plasma can be relevant for the onset of detachment under highly collisional conditions in simple geometries, but play a modest role at low collisionality. Ion‐molecule equipartition presents an additional ion energy loss channel, dissipating ion energy that would otherwise heat the electrons by equipartition. The resulting reduced electron temperature increases the ion‐electron recombination by an order of magnitude as the plasma density is increased, yielding earlier and deeper detachment.

Holm, A.↗
DOE OSTI2019

Modeling K-DEMO Plasma Power Exhaust with UEDGE

The goal of this project is to develop simulation capability for the edge plasma region of K-DEMO in order to assess the heat and particle loads on plasma facing components. The roles of Lawrence Livermore National Laboratory (LLNL) are to provide the UEDGE simulation code for this purpose, to obtain initial edge-plasma solutions for K-DEMO, and to train K-DEMO staff members in both the use of the 2D UEDGE and the code structure such that future modification can be made. A schematic of K-DEMO highlighting the power exhaust is shown in Fig. 1. The major radius of the device is 6.8 m, and the total power to be exhausted is 280 MW in the edge plasma region. Evaluating impact of different divertor geometries and the use of seeded impurity radiation to manage the peak heat flux to divertor-plate surfaces are key aspects that have been addressed during this project.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Calibrated helium and carbon ion flow measurements in the DIII-D divertor plasma

To explore the influence of drifts on ion flows in the DIII-D divertor and Scrape-Off-Layer (SOL), a calibrated Coherence Imaging System (CIS) views the lower divertor. A new tuneable calibration laser and precision wavemeter provide the rest wavelength for CIII 465 nm plasma emissions resulting in ≤ 1 km/s uncertainty in ion velocities that are in the range of 30 km/s in the divertor plasma. The 2-D line-integrated velocity images have been tomographically inverted to obtain maps of ion velocity. Both He + and C 2+ ion velocity maps were measured in both directions of the toroidal magnetic field B T , and C 2+ images were obtained in attached and detached L- and H-mode plasmas. The measured C 2+ velocities are in the range of 20-30km/s and are compared with UEDGE modeling with full drifts. While He II as a main ion is expected to be entrained in the plasma flow, the C 2+ impurity flow involves the competition between the frictional and ion temperature gradient forces. Both the CIS data and UEDGE modeling agree that the C 2+ flow is towards the divertor plates, but flips with the direction of B T . In general, the experimental C 2+ emission and flow profiles have a wider spatial extent than the UEDGE modeling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Tomographic analysis of tangential viewing cameras (invited)

Many tokamaks now use visible light cameras to observe plasma-wall interactions and integrated line emission. The DIII-D coherence imaging spectroscopy diagnostic cameras image interferograms that encode line integrated velocity. By modeling the 2D camera image pixels as line of sight integrals through an axisymmetric discrete grid, it is possible to do tomographic analysis to determine the local plasma line emissivity and parallel velocity. Approaches to solve the inverse problem posed by these tangential viewing cameras are presented. The inversion begins with calculation of the sparse response matrix that encompasses all the geometry and diagnostic information and reduces the process of image formation to a sparse matrix-vector multiply. Our report includes techniques for determining the detailed geometry of the camera views and methods for handling physical quantities that vary spatially. Additionally, the size of the response matrix has driven the development of capability to distribute the coarse parallel calculation across a heterogeneous cluster of computers on the Energy Sciences Network. Iterative techniques are then used to solve the sparse matrix-vector linear system

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

Laser calibration of the DIII-D coherence imaging system

Here, we describe an in situ calibration technique for Coherence Imaging Spectroscopy (CIS) that measures 2-D images of ion flows on DIII-D. A low power CW diode laser that is tuneable in the range 464-468 nm along with a precision wavemeter (10 –5 nm resolution) is used to characterize the interferometer phase as a function of wavelength in the region of C iii (465 nm) and He ii (468 nm). The interferometer is stabilized both mechanically and thermally to minimize drift during the calibration. Optical stirring and an integration sphere are utilized to obtain spatially uniform calibration images. The quality of the calibration data enables a measurement of phase versus wavelength over approximately 10 fringes of the interferometer. These coefficients can also be related to the geometry of the optics and the birefringent crystal of the interferometer. On DIII-D, the integration sphere with the laser light is inserted into the CIS optical system between shots and the laser image and wavelength are automatically recorded, providing a zero velocity reference

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

2D imaging of helium ion velocity in the DIII-D divertor

Two-dimensional imaging of parallel ion velocities are compared to fluid modeling simulations to understand the role of ions in determining divertor conditions and benchmark the UEDGE fluid modeling code. Pure helium discharges are used so that spectroscopic He+measurements represent the main-ion population at small electron temperatures. Electron temperatures and densities in the divertor match simulated values to within about 20-30% establishing the experiment/model match as being at least as good as those normally obtained in the more regularly simulated deuterium plasmas. He + brightness (HeII) comparison indicates that the degree of detachment is captured well by UEDGE, principally due to the inclusion of E x B drifts. Tomographically inverted Coherence Imaging Spectroscopy (CIS) measurements are used to determine the He + parallel velocities which display excellent agreement between model and experiment near the divertor target where He + is predicted to be the main-ion species and where electron-dominated physics dictates the parallel momentum balance. Upstream near the X-point where He + is a minority species and ion-dominated physics plays a more important role, there is an underestimation of the flow velocity magnitude by a factor of 2-3. These results indicate that more effort is required to be able to correctly predict ion momentum in these challenging regimes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Absolute calibration of Doppler coherence imaging velocity images

Here, a new technique has been developed for absolutely calibrating a Doppler Coherence Imaging Spectroscopy interferometer for measuring plasma ion and neutral velocities. An optical model of the interferometer is used to generate zero-velocity reference images for the plasma spectral line of interest from a calibration source some spectral distance away. Validation of this technique using a tunable diode laser demonstrated an accuracy better than 0.2 km/s over an extrapolation range of 3.5 nm; a two order of magnitude improvement over linear approaches. While a well-characterized and very stable interferometer is required, this technique opens up the possibility of calibrated velocity measurements in difficult viewing geometries and for complex spectral line-shapes.

spectroscopy and imaging↗
DOE OSTI2016

Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2 -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2 -seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (see standard divertor).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Asymmetries in the motional Stark effect emission on the DIII-D tokamak [Asymmetries in the motional Stark effect emission]

In this study, spectrometer measurements and filter upgrades to a motional Stark effect polarimeter measuring the outer half-radius of the DIII-D tokamak helped to identify asymmetries in the polarization angle of Stark-split emission. The measured polarization angle of the π components differs and is not orthogonal to the σ component. These differences persist over a range of densities and with low levels of background light. Lastly, it is suggested that the difference in the polarization angle between components is from a change in the ellipticity of the emitted light across the Stark components coupled with imperfect polarization preservation from an in-vessel mirror.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

CORSICA modelling of ITER hybrid operation scenarios

The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor ($q$ ) profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000 s) with a moderate fusion power multiplication factor, $Q$ , of at least 5. This paper presents here candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution]

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower n e , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2 -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2 -seeded SF divertor at P SOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1974

Optimizing subsystem test programs

The goal of reduction of space system costs is pursued in the areas of standardization of subsystem test requirements as well as overall subsystem test program resource requirements. A number of standard components are being developed by NASA. The rationale for the various environmental tests applicable to future missions and launch vehicles is discussed. The need for and approach to developing standard component specifications is presented. In addition, a method for reduction of test expenditures utilizing indexes of mission effect and hardware history as indicators of test requirements is proposed.

Meyer, W. H.↗