Egemen Kolemen
A5033746250- Magnetic confinement fusion research
- Fusion materials and technologies
- Superconducting Materials and Applications
- Ionosphere and magnetosphere dynamics
- Particle accelerators and beam dynamics
- Laser-Plasma Interactions and Diagnostics
- Nuclear reactor physics and engineering
- Plasma Diagnostics and Applications
- Solar and Space Plasma Dynamics
- Astro and Planetary Science
- Spacecraft Dynamics and Control
- Plasma Applications and Diagnostics
- Metallurgical Processes and Thermodynamics
- Nuclear Engineering Thermal-Hydraulics
- Fluid Dynamics and Turbulent Flows
- Spectroscopy and Laser Applications
- Atomic and Subatomic Physics Research
- Advanced Data Storage Technologies
- Laser-induced spectroscopy and plasma
- Gyrotron and Vacuum Electronics Research
- Semiconductor materials and devices
- Spacecraft and Cryogenic Technologies
- Nuclear Physics and Applications
- Stellar, planetary, and galactic studies
- Electrohydrodynamics and Fluid Dynamics
Princeton Plasma Physics Laboratory
2016-2025
Princeton University
2016-2025
Chung-Ang University
2023
The University of Texas at Austin
2022
General Atomics (United States)
2013-2022
Columbia University
2022
Ain Shams University
2022
Oak Ridge National Laboratory
2022
ORCID
2021
Princeton Public Schools
2019
Abstract For stable and efficient fusion energy production using a tokamak reactor, it is essential to maintain high-pressure hydrogenic plasma without disruption. Therefore, necessary actively control the based on observed state, manoeuvre while avoiding tearing instability, leading cause of disruptions. This presents an obstacle-avoidance problem for which artificial intelligence reinforcement learning has recently shown remarkable performance 1–4 . However, obstacle here, difficult...
Abstract The objectives of NSTX-U research are to reinforce the advantages STs while addressing challenges. To extend confinement physics low- A , high beta plasmas lower collisionality levels, understanding transport mechanisms that set performance and pedestal profiles is being advanced through gyrokinetic simulations, reduced model development, comparison NSTX experiment, as well improved simulation RF heating. develop stable non-inductive scenarios needed for steady-state operation,...
Abstract The spherical tokamak (ST) is a leading candidate for Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. National Spherical Torus eXperiment (NSTX) MA-class ST facility in the US actively developing physics basis an ST-based FNSF. In plasma transport research, experiments exhibit strong (nearly inverse) scaling of normalized confinement with collisionality, if this trend holds at low high fusion neutron fluences could be achievable very...
Operating experimental devices have provided key inputs to the design process for ITER axisymmetric control. In particular, experiments quantified controllability and robustness requirements in presence of realistic noise disturbance environments, which are difficult or impossible characterize with modelling simulation alone. This kind information is particularly critical vertical control, poses highest demands on poloidal field system performance, since consequences loss control can be...
Key plasma physics and real-time control elements needed for robustly stable operation of high fusion power discharges in ITER have been demonstrated recent research worldwide.Recent analysis has identified the current density profile as main drive disruptive instabilities simulating ITER's baseline scenario with low external torque.Ongoing development model-based active magnetohydrodynamic is improving stability multiple scenarios.Significant advances made toward physicsbased prediction...
Abstract The ‘Super H-Mode’ regime is predicted to enable pedestal height and fusion performance substantially higher than standard H-Mode operation. This exists due a bifurcation of the pressure, as function density, that by EPED model occur in strongly shaped plasmas above critical density. Experiments on Alcator C-Mod DIII-D have achieved access Super (and Near H) regime, obtained very high including highest tokamak ( p ped ~ 80 kPa) experiments operating near ITER magnetic field. H...
The DESC stellarator optimization code takes advantage of advanced numerical methods to search the full parameter space much faster than conventional tools. Only a single equilibrium solution is needed at each step thanks automatic differentiation, which efficiently provides exact derivative information. A Gauss-Newton trust-region method uses second-order information take large steps in and converges rapidly. With just-in-time compilation GPU portability, high-dimensional runs orders...
We present new stellarator equilibria that have been optimized for reduced turbulent transport using nonlinear gyrokinetic simulations within the optimization loop. The routine involves coupling pseudo-spectral GPU-native code GX with equilibrium and DESC . Since allows fast simulations, we directly optimize heat fluxes. To handle noisy flux traces returned by these employ simultaneous perturbation stochastic approximation (SPSA) method only uses two objective function evaluations a simple...
Abstract The path of tokamak fusion and International thermonuclear experimental reactor (ITER) is maintaining high-performance plasma to produce sufficient power. This effort hindered by the transient energy burst arising from instabilities at boundary plasmas. Conventional 3D magnetic perturbations used suppress these often degrade performance increase risk other instabilities. study presents an innovative field optimization approach that leverages machine learning real-time adaptability...
Steady-state handling of divertor heat flux is a critical issue for ITER and future conventional spherical tokamaks with compact high-power density divertors. A novel ‘snowflake’ (SFD) configuration was theoretically predicted to have significant magnetic geometry benefits mitigation, such as an increased plasma-wetted area higher volume available volumetric power momentum loss processes, compared the standard divertor. Both peak reduction impurity screening been achieved simultaneously core...
Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of heat flux and plate erosion remains to critical issues ITER future concept devices based on conventional spherical geometry with high power density divertors. Experiments conducted in 4–6 MW NBI-heated H-mode plasmas demonstrated is compatible high-confinement core plasma...
Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique operational space is leveraged test theories next-step tokamak operation, including ITER. Present research also examines implications coming device upgrade, NSTX-U. An energy confinement time, τE, scaling unified varied wall conditions exhibits strong improvement of BTτE with decreased electron...
The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in summer of 2015.NSTX-U first plasma subsequently achieved, diagnostic control systems have been commissioned, H-mode accessed, magnetic error fields identified mitigated, physics research campaign carried out.During ten run weeks operation, NSTX-U surpassed record pulse-durations toroidal (TF), high-performance ~1 MA plasmas comparable to best sustained near...
Real-time feedback control based on machine learning algorithms (MLA) was successfully developed and tested DIII-D plasmas to avoid tearing modes disruptions while maximizing the plasma performance, which is measured by normalized beta. The uses MLAs that were trained with ensemble methods using only data available real-time Plasma Control System (PCS) from several thousand discharges. A “tearability” metric quantifies likelihood of onset 2/1 in a given time window, “disruptivity” first...
The extreme heat fluxes in the divertor region of tokamaks may require an alternative to solid plasma-facing components, for extraction and protection surrounding walls. Flowing liquid metals are proposed as alternative, but raise additional challenges that investigation numerical simulations. Free surface designs desirable steady flow profiles stability must be ensured limit undesirable interactions with plasma. Previous studies have mainly used steady-state, 2D, or simplified models...
Abstract For design, scenario planning, and control, ITER all other envisioned tokamaks rely on a variety of statistical physics-based models to extrapolate unseen regimes; most notably from low plasma current high. A "meta-learning" methodology for combining the accuracy data-driven with generalizability is described tested, yielding 5 10 percent improvement in performance beyond either alone task extrapolating time-dependent profile prediction low- high- DIII-D tokamak discharges....
Abstract The Imaging Neutral Particle Analyzer (INPA) at DIII-D is a diagnostic system used to accurately resolve the energy and spatial distributions of fast ions in fusion plasmas. A novel Artificial Intelligence (AI) technique named INPA-net based on Reservoir Computing Networks developed here predict active passive signals produced by charge-exchange reactions from injected edge-cold neutrals, respectively, magnetically confined This model trained using set 21 time domain between 0 s...
<title>Abstract</title> With the increasing urgency of climate change’s impacts and limited success in reducing emissions, “geoengineering,” or artificial manipulation to reduce warming rates, has been proposed as an alternative short-term solution. Options range from taking carbon out atmosphere through sinks brightening clouds planet’s albedo release reflective particles into atmosphere. While still controversial, geoengineering by some a promising low-cost way combating change. In...
Since the first H-mode discharges in 2010, duration of state has been extended and a significantly wider operational window plasma parameters attained. Using second neutral beam (NB) source improved tuning equilibrium configuration with real-time control, stored energy Wtot ∼ 450 kJ achieved corresponding confinement time τE 163 ms. Recent discharges, produced fall 2012, have reached βN up to 2.9 surpassed n = 1 ideal no-wall stability limit computed for pressure profiles, which is one key...
ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, limited number of operational discharges. In addition, high energy content, heat fluxes, neutron very long pulse operation place novel demands on in many areas ranging from boundary divertor regulation to kinetics stability control. Both experimental operations schedules provide time for tuning algorithms relative operating devices. Although...
Real-time steerable electron cyclotron current drive (ECCD) has been demonstrated to reduce the power requirements and time needed remove 3/2 2/1 neoclassical tearing modes (NTMs) in DIII-D tokamak. In a world first demonstration of techniques required ITER, island formation onset is detected automatically, gyrotrons are turned on real-time ECCD launcher mirrors moved promptly at location islands. This shrinks suppresses well before saturation using motional Stark effect constrained...
Liquid metal (LM) plasma-facing components (PFCs) may provide a resolution to the challenging fusion environment, particularly first wall and divertor surfaces. Transforming these concepts into viable technologies will require considerable research development. With nuclear regime in mind, Fusion Energy System Studies group examined LM PFCs order identify needed thrusts that could accelerate their development assess viability. behavior, solid substrate aspects, facility integration aspects...