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Canik, J. M.

Publications and source records attributed to Canik, J. M..

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Reversed-direction 2-point modelling applied to divertor conditions in DIII-D *

Abstract A predictive form of the extended 2-point model known as the ‘reverse 2-point model’, Rev2PM, is applied to a range of detachment levels in the open lower divertor of DIII-D, showing that the experimentally measured electron temperature ( T e ) and pressure ( p e ) at the divertor entrance can be calculated within 50% from target measurements, if and only if a posteriori corrections for convective heat flux are included in the model. Unlike the standard 2-point model, the Rev2PM calculates upstream scrape-off layer (SOL) quantities (such as separatrix T e and p e ) from target conditions (such as T e and parallel heat flux), with volumetric power and momentum losses depending solely on target T e . The Rev2PM is tested against a database of DIII-D inter-ELM divertor Thomson scattering measurements, built from a series of 6 MW, 1.3 MA, LSN H-mode discharges with varied main ion density, drift direction, and nitrogen puffing rate. Measured target T e ranged from 0.4–25 eV over this database, and upstream T e ranged from 5–60 eV. Poor agreement is found between upstream measurements and Rev2PM calculations that assume purely conductive parallel heat transport. However, introducing a posteriori corrections to account for convective heat transport brings the Rev2PM calculations within 50% of the measured upstream values across the dataset. These corrections imply that up to 99% of the parallel heat flux is carried by convection in detached conditions in the DIII-D open lower divertor, though further work is required to assess any potential dependencies on device size or divertor closure.

2-point model↗
DOE OSTI2022

A reduced model for the ITER divertor based on SOLPS solutions for ITER Q = 10 baseline conditions: A. identifying options for the control parameters*

Edge codes such as SOLPS coupled to neutral codes such as EIRENE have become so comprehensive and sophisticated that they now constitute, in effect, 'code-experiments' that, as for actual experiments, can benefit from interpretation using simple models and conceptual frameworks, i.e. reduced models. The first task is the identification of options for the reduced model control parameters that are best suited for control of the action of the divertor, i.e. for control of target power loading and sputter-erosion, primarily. A strong correlation between the electron temperature at the divertor target, T e,t , and the neutral deuterium D 2 density at the target, n D2,t , flux-tube resolved, has recently been reported for a number of code studies including SOLPS-4.3 modeling of a set of ~50 ITER baseline cases: Q DT = 10, q 95 = 3, P SOL = 100 MW, metallic walls, and Ne seeding. This part A of the present study reports new results for largely the same ITER cases, confirming the strong correlations reported earlier between local values of T e,t , and (i) n D2,t , and (ii) normalized volumetric losses of power and pressure in the divertor. Strong correlations have now also been found, and are reported here for the first time, between T e,t and all of the divertor target quantities of practical interest. A physical explanation for this surprising result has not been fully identified; nevertheless it has encouraging implications for reduced modeling of the ITER divertor. For such ITER conditions, (i) the global Ne injection rate, Inj Ne [Ne/s], and (ii) the electron temperature at the location on the target where the peak power deposition occurs, T e,t @q ⊥,pk (eV), are found to be promising reduced model control parameters. In the companion report, part B, a reduced model for the ITER divertor is developed and described in detail, based on reversed-direction 2 point modelling, Rev2PM. The input to the reduced model is a value of the variable pair (T et @q ⟂,pk ,Inj Ne ) and the output are values of the various target as well as divertor-entrance quantities of practical interest, e.g. q ⊥,pk , n e,Xpt (the electron density at the poloidal location of the X-point), etc. In part B the reduced model is quantitatively characterized using one half of the code cases; it is then used to successfully predict (replicate) the code values of e.g. n e,Xpt for the other half of the cases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Quantification of the effect of uncertainty on impurity migration in PISCES-A simulated with GITR

A Bayesian inference strategy has been used to estimate uncertain inputs to global impurity transport code (GITR) modeling predictions of tungsten erosion and migration in the linear plasma device, PISCES-A. This allows quantification of GITR output uncertainty based on the uncertainties in measured PISCES-A plasma electron density and temperature profiles (n e , T e ) used as inputs to GITR. The technique has been applied for comparison to dedicated experiments performed for high (4 × 10 22 m –2 s –1 ) and low (5 × 10 21 m –2 s –1 ) flux 250 eV He–plasma exposed tungsten (W) targets designed to assess the net and gross erosion of tungsten, and corresponding W impurity transport. The W target design and orientation, impurity collector, and diagnostics, have been designed to eliminate complexities associated with tokamak divertor plasma exposures (inclined target, mixed plasma species, re-erosion, etc) to benchmark results against the trace impurity transport model simulated by GITR. The simulated results of the erosion, migration, and re-deposition of W during the experiment from the GITR code coupled to materials response models are presented. Specifically, the modeled and experimental W I emission spectroscopy data for a 429.4 nm line and net erosion through the target and collector mass difference measurements are compared. Furthermore, the methodology provides predictions of observable quantities of interest with quantified uncertainty, allowing estimation of moments, together with the sensitivities to plasma temperature and density.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Inter-ELM pedestal evolution and the role of edge fluctuations in the C-Mod and DIII-D tokamaks

Edge localized modes (ELMs) lead to a cyclical behaviour of the H-mode pedestal density, temperature and pressure. Substantial evidence exists associating the ELM crash with violation of ideal MHD stability, via current driven kink/peeling modes and/or pressure gradient driven ballooning modes. Recovery from ELM crashes and buildup to the next ELM crash is an active and critical area of edge physics research for ITER projections. In one set of theories, the pedestal pressure profile in H-mode discharges is predicted to be limited by micro-instabilities. One such micro-instability is the kinetic ballooning mode (KBM), which is hypothesized to provide a ‘soft’ limit that regulates the edge transport, thereby restricting the local pressure gradient. In this picture, e.g. in the EPED model, the pedestal width expands slowly until the ideal MHD stability limit is reached. Experiments were recently performed on the C-Mod and DIII-D devices to search for instabilities correlated with the pedestal evolution between ELMs. The results show correlations between the onset of quasi-coherent fluctuations between type I ELMs and a critical temperature gradient. Linear gyrokinetic calculations have shown that these fluctuations have characteristics similar to those expected for KBMs. These results provide additional data toward validation of the EPED model, increasing confidence in ITER projections.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Testing predictions of electron scale turbulent pedestal transport in two DIII-D ELMy H-modes

In this study, we present linear and nonlinear gyrokinetic analyses in the pedestal region of two DIII-D ELMy H-mode discharges using the CGYRO code. The otherwise matched discharges employ different divertor configurations to investigate the impact of varying recycling and particle source on pedestal profiles. Linear gyrokinetic simulations find electrostatic ion-scale instabilities (ion temperature gradient and trapped electron modes, ITG-TEM) are present just inside the top of the pedestal with growth rates that are enhanced significantly by parallel velocity shear. In the sharp gradient region, E x B shearing rates are comparable or larger than ion scale growth rates, suggesting the suppression of ITG-TEM modes in this region. Instead, the electron temperature profiles are found to be correlated with and just above the electron temperature gradient (ETG) instability thresholds. Using gradients varied within experimental uncertainties, nonlinear electron-scale gyrokinetic simulations predict electron heat fluxes from ETG turbulence, that when added to neoclassical ion thermal transport simulated by NEO, account for 30-60% of the total experimental heat flux. In addition, the neoclassical electron particle flux is found to contribute significantly to the experimental fluxes inferred from SOLPS-ITER analysis. Additional nonlinear gyrokinetic simulations are run varying input gradients to develop a threshold-based reduced model for ETG transport, finding a relatively simple dependence on η e = L ne /L Te . Predictive transport simulations are used to validate this pedestal-specific ETG model, in conjunction with a model for neoclassical particle transport. In both discharges, the predicted electron temperatures are always overpredicted, indicative of the insufficient stiffness in the ETG pedestal model to account for all of the experimental electron thermal transport. In the case of the closed divertor discharge with lower particle source, the predicted electron density is close to the experiment, consistent with the magnitude of neoclassical particle transport in that discharge. However, the density profiles are overpredicted in the open divertor discharge (larger particle source), due to insufficient model transport. In conclusion, the implications for other mechanisms accounting for the remainder of transport in the sharp gradient region in the two discharges are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Optimization of pumping performance in the EAST upgraded divertor

The pumping performance of the EAST tokamak upgraded divertor has been optimized by varying the height and width of the duct that connects the plasma and plenum volumes. The optimization is enacted using a semi-analytic model which predicts the molecular pressure in the plenum given the plasma facing component geometry and plasma parameters along the divertor, specifically electron temperature, electron density, and the ion particle flux. The model is semi-analytic as the plasma parameters can come from experimental data (or plasma transport simulations), and uses a first-flight approximation for neutral transport. This model is computationally inexpensive, allowing for several geometries and sets of plasma parameters to be rapidly evaluated. A duct geometry is discovered that performs well for different plasma equilibria and strike point positions. The semi-analytic model results are compared to two dimensional plasma and neutral transport simulations using the SOLPS-ITER code, suggesting that charge-exchange is a significant contributor to the plenum pressure at low electron temperature. An extension to the semi-analytic model improves the comparison over a broader range of temperature. The conservative prediction of the model and the broad maximum in the pressure contours imply that the inclusion of charge-exchange is not required in the optimization procedure, but that quantitative estimates should be provided with a high-fidelity model such as SOLPS-ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Active Recycling Control Through Lithium Injection in EAST

The coating of tokamak walls with thin layers of lithium has been demonstrated to reduce plasma recycling from the plasma-facing surfaces and to improve overall plasma performance. These effects, including reduced divertor D α emission, the elimination of edge-localized modes, and increased energy confinement have been observed in multiple experiments when lithium coatings are applied before plasma discharges. However, this coating technology does not extrapolate to future long-pulse devices, since the lithium coatings will be passivated by the continual plasma flux onto the surface. In order to provide active conditioning capability, a new technology has been developed that is capable of injecting lithium powder into the scrape-off layer plasma during plasma discharges, where it quickly liquefies and turns into an aerosol. The use of this “lithium dropper” is under study at the Experimental Advanced Superconducting Tokamak (EAST), where the potential benefits of real-time wall conditioning via lithium injection are being tested. Here, we present an analysis of the recycling characteristics during EAST experiments testing active lithium injection in order to assess recycling reduction and control. Lithium aerosol was injected from the top of the machine, with one system dropping lithium near the X-point and another into the low-field side divertor leg. The injection of lithium into the SOL reduced divertor recycling, as evidenced by reduced D α emission with ion flux measured by probes relatively unchanged. This effect is strongest in the active divertor, confirming the lithium is transported to strongly plasma-wetted areas. Furthermore, quantitative analysis of the recycling changes using the SOLPS edge plasma and neutral transport code indicated a ~20% reduction in recycling coefficient with lithium injection.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

Here, we report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previous ELM elimination results via Li injection into the lower carbon divertor in EAST. These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs, highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Developing the science and technology for the Material Plasma Exposure eXperiment

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1–10 eV and electron densities of $10^{21}{\text{}}\!-\!10^{20}$ ${\rm m}^{-3}$. The resulting heat fluxes are about 10 MW ${\rm m}^{-2}$ . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW ${\rm m}^{-2}$ was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to $1.1 \times 10^{20}$ ${\rm m}^{-3}$ at high magnetic fields of 1.0 T at the target. Finally, the experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL's High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Compatibility of lithium plasma-facing surfaces with high edge temperatures in the Lithium Tokamak Experiment

We measured high edge electron temperatures (200 eV or greater) at the wall-limited plasma boundary in the Lithium Tokamak Experiment (LTX). Flat electron temperature profiles are a long-predicted consequence of low recycling boundary conditions. Plasma density in the outer scrape-off layer is very low, 2-3 x 10(17) m(-3), consistent with a low recycling metallic lithium boundary. In spite of the high edge temperature, the core impurity content is low. Z(eff) is estimated to be similar to 1.2, with a very modest contribution (< 0.1) from lithium. Experiments are transient. Gas puffing is used to increase the plasma density. After gas injection stops, the discharge density is allowed to drop, and the edge is pumped by the low recycling lithium wall. An upgrade to LTX-LTX-beta, which includes a 35A, 20 kV neutral beam injector (on loan to LTX from Tri-Alpha Energy) to provide core fueling to maintain constant density, as well as auxiliary heating, is underway. LTX-beta is briefly described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Preliminary design of a tangentially viewing imaging bolometer for NSTX-U

The infrared imaging video bolometer (IRVB) measures plasma radiated power images using a thin metal foil. Two different designs with a tangential view of NSTX-U are made assuming a 640 × 480 (1280 × 1024) pixel, 30 (105) fps, 50 (20) mK, IR camera imaging the 9 cm × 9 cm × 2 μm Pt foil. The foil is divided into 40 × 40 (64 × 64) IRVB channels. This gives a spatial resolution of 3.4 (2.2) cm on the machine mid-plane. The noise equivalent power density of the IRVB is given as 113 (46) μW/cm{sup 2} for a time resolution of 33 (20) ms. Synthetic images derived from Scrape Off Layer Plasma Simulation data using the IRVB geometry show peak signal levels ranging from ∼0.8 to ∼80 (∼0.36 to ∼26) mW/cm{sup 2}.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Evidence of toroidally localized turbulence with applied 3D fields in the DIII-D tokamak

New evidence indicates that there is significant 3D variation in density fluctuations near the boundary of weakly 3D tokamak plasmas when resonant magnetic perturbations are applied to suppress transient edge instabilities. The increase in fluctuations is concomitant with an increase in the measured density gradient, suggesting that this toroidally localized gradient increase could be a mechanism for turbulence destabilization in localized flux tubes. Two-fluid magnetohydrodynamic simulations find that, although changes to the magnetic field topology are small, there is a significant 3D variation of the density gradient within the flux surfaces that is extended along field lines. This modeling agrees qualitatively with the measurements. The observed gradient and fluctuation asymmetries are proposed as a mechanism by which global profile gradients in the pedestal could be relaxed due to a local change in the 3D equilibrium. In conclusion, these processes may play an important role in pedestal and scrape-off layer transport in ITER and other future tokamak devices with small applied 3D fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Effect of progressively increasing lithium conditioning on edge transport and stability in high triangularity NSTX H-modes

A sequence of H-mode discharges with increasing levels of pre-discharge lithium evaporation (‘dose’) was conducted in high triangularity and elongation boundary shape in NSTX. Energy confinement increased, and recycling decreased with increasing lithium dose, similar to a previous lithium dose scan in medium triangularity and elongation plasmas. Data-constrained SOLPS interpretive modeling quantified the edge transport change: the electron particle diffusivity decreased by 10-30x. The electron thermal diffusivity decreased by 4x just inside the top of the pedestal, but increased by up to 5x very near the separatrix. These results provide a baseline expectation for lithium benefits in NSTX-U, which is optimized for a boundary shape similar to the one in this experiment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Linear gyrokinetic simulations of microinstabilities within the pedestal region of H-mode NSTX discharges in a highly shaped geometry

Linear (local) gyrokinetic predictions of edge microinstabilities in highly shaped, lithiated and non-lithiated NSTX discharges are reported using the gyrokinetic code GS2. Microtearing modes dominate the non-lithiated pedestal top. The stabilization of these modes at the lithiated pedestal top enables the electron temperature pedestal to extend further inwards, as observed experimentally. Kinetic ballooning modes are found to be unstable mainly at the mid-pedestal of both types of discharges, with un- stable trapped electron modes nearer the separatrix region. At electron wavelengths, ETG modes are found to be unstable from mid-pedestal outwards for η e, exp ~2.2 with higher growth rates for the lithiated discharge. Near the separatrix, the critical temperature gradient for driving ETG modes is reduced in the presence of lithium, re ecting the reduction of the lithiated density gradients observed experimentally. A preliminary linear study in the edge of non-lithiated discharges shows that the equilibrium shaping alters the electrostatic modes stability, found more unstable at high plasma shaping.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Investigation of island formation due to RMPs in DIII-D plasmas with the SIESTA resistive MHD equilibrium code

The SIESTA magnetohydrodynamic (MHD) equilibrium code has been used to compute a sequence of ideally stable equilibria resulting from numerical variation of the helical resonant magnetic perturbation (RMP) applied to an axisymmetric DIII-D plasma equilibrium. Increasing the perturbation strength at the dominant m=2, n=-1 , resonant surface leads to lower MHD energies and increases in the equilibrium island widths at the m=2 (and sidebands) surfaces, in agreement with theoretical expectations. Island overlap at large perturbation strengths leads to stochastic magnetic fields which correlate well with the experimentally inferred field structure. The magnitude and spatial phase (around the dominant rational surfaces) of the resonant (shielding) component of the parallel current are shown to change qualitatively with the magnetic island topology.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗