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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Understanding JET-C quiescent phases with edge harmonic magnetohydrodynamic activity and comparison with behaviour under ITER-like wall conditioning

An analysis of edge localised mode-free (quiescent) H-mode discharges exhibiting edge harmonic magnetoydrodynamic activity in the JET-carbon wall machine is presented. It is observed that the otherwise quiescent pulses with multiple-n harmonic oscillations are sustained until a threshold in pedestal electron density and collisionality is crossed. The macroscopic pedestal parameters associated with the quiescent phase are compared with those of a database of JET-ELMy discharges with both carbon and ITER-like wall (ILW). This comparison provides the identification of the existence regions in the relevant pedestal and global plasma parameters for edge harmonic oscillations (EHOs) in JET plasmas. Although the ELMy database scans pedestal collisionality and β values typical of ET-carbon quiescent operation, shaping and current are not simultaneously compatible with EHO existence. Nevertheless, ILW operation with JET-carbon quiescent-like parameters could in principle be achieved, and improved pedestal performance could be observed in more recent JET-ILW pulses.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Physics-Based Real-time Analysis and Control to Achieve Transient-Free Operations for the ITER Era (Final Report)

This is the final report for the early career award for Prof. Egemen Kolemen that aimed to develop the physics and technical basis of real-time transient – i.e. disruptions and Edge Localized Modes (ELMs) – avoidance and feedback control at the DIII-D nuclear fusion facility for the ITER era. During this project, we reconstructed high quality real-time DIII-D equilibria using kinetic information from the motional Stark effect (MSE), Thomson Scattering (TS), and Charge Exchange Recombination (CER) spectroscopy. These equilibria were analyzed in real-time to determine the low-n ideal MHD stability and the pedestal stability thresholds, and to obtain plasma response models. These calculations were used to understand which physical process evolves to lead to the onset of disruption in ITER-relevant scenarios at DIII-D, and how disruption can be predicted using stability analysis as opposed to the current methods of correlation of various signals, which are descriptive but lack predictive power. In the disruption studies, we focused on tearing modes, since these constitute the vast majority of disruptions in this regime. With this knowledge, real-time disruption avoidance algorithms were developed and tested. The enhanced situation awareness and physics insight were used in advanced real-time feedback control of pedestal and core profiles, which allowed the pursuit and sustainment of transient-free high-confinement regimes that cannot be achieved via feed-forward design based on in-between shot analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

The energy confinement response of DIII-D plasmas to Resonant Magnetic Perturbations

Here, Resonant Magnetic Perturbations (RMPs) are a leading method for edge localized modes (ELMs) Control in fusion plasmas. However they can also cause a rapid degradation in energy confinement. In this paper we show that the energy confinement in low collisionality (v* e < 0.3) DIII-D ITER Similar Shape (ISS) plasmas often recovers after several energy confinement times for RMP amplitudes up to the threshold for ELM suppression. Immediately following the application of the RMP, the plasma stored energy decreases in proportion to the decrease in the line-averaged density during density "pump-out". Later in the discharge confinement recovery is observed in the thermal ion channel and is correlated with the increase in the ion temperature at the top of the H-mode pedestal. A correlation between the inverse scale length of the ion temperature (α/L Ti ) and the E x B shearing rate at the top of the pedestal is seen during the confinement recovery phase. Transport analysis reveals that the confinement improvement in the ion channel results from the self-similarity in the ion temperature profiles in the plasma core combined with the observed increase in α/L Ti in the plasma edge following density pump-out. In contrast the electron temperature scale length (α/L Ti ) remains essentially unchanged in response to the application of the RMP. At significantly higher RMP levels the edge EXB shearing rate and α/L Ti does not increase and the confinement does not recover following density pump-out.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

SU-E-J-261: The Importance of Appropriate Image Preprocessing to Augment the Information of Radiomics Image Features

Purpose: To investigate how different image preprocessing techniques, their parameters, and the different boundary handling techniques can augment the information of features and improve feature’s differentiating capability. Methods: Twenty-seven NSCLC patients with a solid tumor volume and no visually obvious necrotic regions in the simulation CT images were identified. Fourteen of these patients had a necrotic region visible in their pre-treatment PET images (necrosis group), and thirteen had no visible necrotic region in the pre-treatment PET images (non-necrosis group). We investigated how image preprocessing can impact the ability of radiomics image features extracted from the CT to differentiate between two groups. It is expected the histogram in the necrosis group is more negatively skewed, and the uniformity from the necrosis group is less. Therefore, we analyzed two first order features, skewness and uniformity, on the image inside the GTV in the intensity range [−20HU, 180HU] under the combination of several image preprocessing techniques: (1) applying the isotropic Gaussian or anisotropic diffusion smoothing filter with a range of parameter(Gaussian smoothing: size=11, sigma=0:0.1:2.3; anisotropic smoothing: iteration=4, kappa=0:10:110); (2) applying the boundaryadapted Laplacian filter; and (3) applying the adaptive upper threshold for the intensity range. A 2-tailed T-test was used to evaluate the differentiating capability of CT features on pre-treatment PT necrosis. Result: Without any preprocessing, no differences in either skewness or uniformity were observed between two groups. After applying appropriate Gaussian filters (sigma>=1.3) or anisotropic filters(kappa >=60) with the adaptive upper threshold, skewness was significantly more negative in the necrosis group(p<0.05). By applying the boundary-adapted Laplacian filtering after the appropriate Gaussian filters (0.5 <=sigma<=1.1) or anisotropic filters(20<=kappa <=50), the uniformity was significantly lower in the necrosis group (p<0.05). Conclusion: Appropriate selection of image preprocessing techniques allows radiomics features to extract more useful information and thereby improve prediction models based on these features.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

Final Progress Report The U.S. Department of Energy Research Grant No. DE-SC0008660 Plasma Surface Interactions: Bridging from the Surface to the Micron Frontier through Leadership Class Computing

The choice of material for the plasma facing components (PFC), in particular, for divertor targets, is one of the main issues for future tokamak reactors. There are two major requirements for the PFC’s material: acceptable level of tritium retention and durability in a harsh environment of fusion grade plasma. Based on these criteria, some years ago it was decided that tungsten is an acceptable material for divertor targets in ITER. However, further experimental studies reveal that the irradiation of tungsten even with low energetic (well below sputtering threshold!) He containing plasma causes significant modification of surface morphology, formation of the layer of He nano-bubbles (in the temperature range T<1000 K), “fuzz” (for 1000 K 2000 K) (e.g. see Fig. 1). Recall that He, being an ash of D-T fusion reactions, is an inherent impurity in fusion plasma. The goals of the UCSD Applied Plasma Theory Group was: i) investigate the mechanisms of the formation of He nano-bubble layer and fuzz growth under He irradiation, as well as the physics of transport of hydrogen species in tungsten lattice, and ii) develop physics understanding of the models suitable for the incorporation into the Xolotl-PSI code based on the reaction-diffusion approach, which is the flagship of the whole SciDAC project [8], which can guide both numerical simulations and experimental studies. Here we just highlight our major accomplishments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2021

Generation-based Evolutionary Tool for the Optimization of Constellations (GenETOC)

With the rapid growth in the capabilities of smaller satellites, satellite architectures that replace a single, extremely capable spacecraft with multiple, cheaper ones are gaining in popularity. Unfortunately, the orbit design process for constellations can be significantly more involved, especiallywhen the relative placement of the individual spacecraft within the constellation is not constrained by mission and/or science objectives. Optimizing a satellite constellation in the presence of multiple, competing objectives is a highly complex problem to which many traditional mathematical optimization methods cannot be applied and few tools exist to help mission designers search for promising candidate mission designs. The Generation-based Evolutionary Tool for the Optimization of Constellations (GenETOC) has been created to search for near-optimal constellation design options. GenETOC combines a modified version of the Non-dominated Sorting Genetic Algorithm II (NSGA II) with STK Components libraries (a 3rdparty .NET package created by Analytical Graphics Inc.) to create a framework that enables a mission designer to generate a simulation that models the design problem and obtain a family of potential, near-optimal solutions that can be investigated more in detail. GenETOC was developed in C# using the .NET framework with Windows Presentation Foundation (WPF) serving as the framework from which to create the graphical user interface (GUI). GenETOC user inputs can be categorized into three major data components: definition of the problem (areas of interest, satellite decision parameters, and sensor configurations), definition of performance objectives, and specification of the genetic algorithm (GA) parameters. In the problem definition component, the user is prompted to define the areas of interest against which the performance metrics will be computed, define the sensor parameters and attach them to specific spacecraft, select which satellite orbital parameters will be added to the decision space of the GA, and specify the range of desired values for each optimization parameter. For performance objectives, the user is presented with a list of available coverage and revisit performance based calculation options from which two metrics are chosen to serve as the objective functions that the GA will use to evaluate solutions during the optimization process. Finally, the definition of the GA parameters provides user control over the number of generations (number of optimization iterations), the population size (number of candidate constellations created in each generation), and the adaptive mutation and crossover threshold values (control parameters for how frequently each process occurs during the optimization). GenETOC has been extensively tested to verify the individual components of the optimization process. The GA has been tested against a suite of GA test problems to confirm convergence to the known two and three-dimensional Pareto fronts. The coverage and revisit performance metrics obtained in GenETOC are compared with STK desktop scenarios, confirming the constellations are being appropriately modeled within GenETOC simulations. A walkthrough of a simple, example problem is provided to illustrate the workings of GenETOC and to demonstrate the output available to the mission designer.

mission design↗
DOE OSTI2016

An n -material thresholding method for improving integerness of solutions in topology optimization

It is common in solving topology optimization problems to replace an integer-valued characteristic function design field with the material volume fraction field, a real-valued approximation of the design field that permits "fictitious" mixtures of materials during intermediate iterations in the optimization process. This is reasonable so long as one can interpolate properties for such materials and so long as the final design is integer valued. For this purpose, we present a method for smoothly thresholding the volume fractions of an arbitrary number of material phases which specify the design. This method is trivial for two-material design problems, for example, the canonical topology design problem of specifying the presence or absence of a single material within a domain, but it becomes more complex when three or more materials are used, as often occurs in material design problems. We take advantage of the similarity in properties between the volume fractions and the barycentric coordinates on a simplex to derive a thresholding, method which is applicable to an arbitrary number of materials. As we show in a sensitivity analysis, this method has smooth derivatives, allowing it to be used in gradient-based optimization algorithms. Finally, we present results, which show synergistic effects when used with Solid Isotropic Material with Penalty and Rational Approximation of Material Properties material interpolation functions, popular methods of ensuring integerness of solutions.

42 ENGINEERING↗
DOE OSTI2021

Characteristics of Ultralow Aspect Ratio Toroidal Plasmas and Non-Solenoidal Startup and Edge Stability Studies at Near-Unity Aspect Ratio (Final Report)

This grant supported a long-term research and development program to explore the physics and technology of spherical tokamak (ST) plasma behavior at near-unity aspect ratio ( A ). The PEGASUS Toroidal Experiment is a university-scale ( R ) ~ 0.40 m, α ~ .35 m, I p ≤ 0.3 MA, B T ≤ 0.15 T) experiment that accesses A → 1, to provide large values of the normalized plasma current, I N = I p /α B T , or equivalently the toroidal field utilization factor I p / I TF . The research activities focused on: 1) developing the technique of Local Helicity Injection (LHI) as a viable means of initiating and growing an ST plasma without the use of a central solenoid; 2) exploring the equilibrium and stability properties of spherical tokamak plasmas at extremely low aspect ratio; and 3) obtaining new insights in the properties of the enhanced confinement H-mode regime at A ~ 1. The initial (Phase I) facility employed a unique high-stress central solenoid Ohmic for current drive and heating with simple magnet current waveforms. A subsequent major upgrade (Phase II) offered more flexible control of the plasma evolution along with a powerful LHI system for plasma initiation, heating and sustainment. The LHI technique consists of injection of high-density current streams into vacuum toroidal and vertical magnetic fields. These streams become unstable and support magnetic reconnection to allow the system to relax into a tokamak-like magnetic configuration. This program focused on: 1) development of electron current sources that provide sufficiently high injection currents in the challenging edge plasma region; 2) validation of theoretical limits to achievable I p arising from magnetic helicity and energy conservation plus absolute limits arising from relaxation to a minimal energy states; 3) tests of a power balance model to describe the evolution of the plasma current I p (t); and 4) study of magnetic fluctuations and their role in the I p evolution. This culminated in the production of toroidal currents of I p ≥ 0.2 MA with only ~8 kA of injected current. Operation with Ohmic induction alone in Phase I created plasmas with soft limits on the achievable I p and field utilization ( I p / I TF ~ 1) due to the onset of large low-order (2/1 or 3/2) internal resonant tearing modes. With the expanded capabilities in Phase II operation, methods employed to mitigate those modes and achieve I p / I TF > 1 were: 1) I TF ramp-downs; 2) non-solenoidal startup via LHI; and 3) extremely low I TF Ohmic operations assisted by LHI current injector pre-ionization. Extensive use of LHI-driven startup and drive for non-solenoidal operation produced favorable current profiles to stably operate tokamak plasmas with unprecedentedly high I N = I p / aB T ~ 15 and I p / I TF > 2. With strong reconnection-driven anomalous ion heating, this in turn provided access to tokamak plasmas with B t up to 100%, a world record 2 times higher than that previously achieved in advanced or spherical tokamaks. This high B t regime had a large fraction (~50%) of the plasma volume in a region of an absolute minimum- B configuration. The MHD stability limit in this regime was the ideal external kink mode without a stabilizing wall. These results provided the first demonstration of access to tokamak plasmas with very high B t , high i N , high elongation κ, and low internal inductance ℓ i , and as such served as a proof-ofprinciple for access to this long-sought ST regime. The naturally low B T needed for stability at A ≤ 1.2 reduces the L-H mode power threshold, P LH , so that H-mode plasmas were achieved with Ohmic heating alone. The power threshold for the transition to the H-mode regime was found to be 10× the accepted values from the ITER international scaling, confirming initial trends indicated by experiments on MAST and NSTX, and suggesting that the physics behind this transition is not yet fully understood. The density dependence of the transition and the observed insensitivity to whether the plasma is bounded by a limiter or magnetic divertor are consistent with the recent FM 3 model. Multiple unstable toroidal magnetic modes were measured during ELM (Edge Localized Mode) crashes, with generally lower toroidal mode numbers than seen at higher A , which was attributed to the larger peeling mode instability drive at low A . First-ever edge current profile measurements through an ELM crash showed current-carrying filaments were generated, as seen in ELM simulations and in electromagnetic blob transport theory.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Machine Learning for Real-time Fusion Plasma Behavior Prediction and Manipulation (Final Report)

The goal of this project is to implement real-time analysis of 2D Beam Emission Spectroscopy (BES) data to predict and control transient and high-bandwidth events at DIII-D. In essence, we wish to bring high-bandwidth fluctuation diagnostics into the realm of real-time measurements and control. The BES ML models will necessarily be deep neural networks (DNN) with a “data flow” architecture for compatibility with high-throughput, low-latency evaluation on a field-programmable gate array (FPGA) or other emerging processor technologies. The real-time output will be fed to the plasma control system (PCS) for real-time control tasks, specifically for ELM control and avoidance and for QH-mode access and sustainment. We anticipate that the real-time analysis of fluctuation diagnostics will create new enabling technologies to predict and control transient events such as confinement mode transitions, edge-localized modes, Alfven eigenmode events, and disruptions. The proposed research is aligned with ITER research needs and DIII-D programmatic goals. For instance, the prediction and avoidance of ELM events is critical for ITER machine safety. Also, H-mode access with RMP ELM suppression in ITER is an active research area due to high separatrix density, narrow SOL width, and elevated LH transition power threshold.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Development of robust and multi-mode control of tearing in DIII-D

Neoclassical tearing modes (NTMs) are instabilities that can produce undesirable magnetic islands in tokamak plasmas. They can be stabilized by applying electron cyclotron current drive (ECCD) at the island. The NTM control system on DIII-D can now control multiple modes. Each of 6 mirrors that reflect ECCD beams into the plasma can be assigned to different surfaces in the plasma where NTMs are unstable. The control system then steers the mirrors to keep the beams aimed at the surfaces. The system routinely stabilizes one NTM preemptively and has now also been used to control two modes in the same discharge. With the “catch-and-subdue” function, ECCD-generating gyrotrons can be turned on when NTMs appear and off after suppression. Newly triggered NTMs can be promptly suppressed if mode onset is detected early and ECCD immediately applied. Early mode detection is achieved in this paper by spectral analysis of Mirnov probes with a band-pass filter for the expected mode frequency. Targeted surfaces are tracked by equilibrium reconstructions (that include measurements of the motional Stark effect). The ECCD position is tracked by ray-tracing using the TORBEAM code. Several techniques are being explored for fine-tuning alignment when NTMs occur. One method adjusts ECCD alignment in steps until the island decays fast enough. A second method sweeps the alignment to find the optimum. A third method pulses gyrotrons and uses electron cyclotron emission to compare where the resulting temperature pulses are relative to temperature fluctuations from a rotating NTM. NTM control in ITER is expected to use active profile regulation to maximize controllability, followed by repeated catch-and-subdue actions if modes are retriggered, in order to maintain island size below the disruptive threshold while maximizing confinement and fusion gain. Between events, real-time tracking will be performed to maintain alignment and readiness for subsequent catch-andsubdue actions. Methods for active probing of stability boundaries will be studied as possible diagnostics for the profile regulation. Finally, selected elements of this ITER NTM control vision will be discussed and assessed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Possible influence of near SOL plasma on the H-mode power threshold

A strong effect of divertor configuration on the threshold power for the L-H transition (P LH ) was observed in recent JET experiments in the new ITER-like Wall (ILW). Following a series of EDGE2D-EIRENE code simulations with Be impurity and drifts a possible mechanism for the P LH variation with the divertor geometry is proposed. Both experiment and code simulations show that in the configuration with lower neutral recycling near the outer strike point (OSP), electron temperature (T e ) peaks near the OSP prior to the L-H transition, while in the configuration with higher OSP recycling T e peaks further out in the scrape-off layer (SOL) and the plasma stays in the L-mode at the same input power. Code results show large positive radial electric field (E r ) in the near SOL under lower recycling conditions leading to a large E × B shear across the separatrix which may trigger earlier (at lower input power) edge turbulence suppression and lower P LH . Suppressed T e ‘s at OSP in configurations with strike points on vertical targets (VT) were observed earlier and explained by a geometrical effect of neutral recycling near this particular position, whereas in configurations with strike points on horizontal targets (HT) the OSP appears to be more open for neutrals.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

X-ray diagnostic developments in the perspective of DEMO

Soft X-ray diagnostics at present are not adequate for a burning plasma experiment, neither in term of hardware nor as diagnostic conception. Detectors have to be radiation tolerant, easily shielded, with low sensitivity to neutrons and gammas and with energy discrimination. Layout and viewing capability should be more flexible, thanks to the use also of optical devices, going toward a configuration intermediate between discrete tomography and pure imaging. The general conception of these diagnostics should therefore evolve in the direction of pattern recognition for a real time feedback. This work is focused on the diagnostic developments undertaken at the ENEA- Frascati X-ray Laboratory, following in particular three directions: gas detector for fast and advanced high density tomography, C-MOS solid state imaging detectors for slow control and X-ray polycapillary optics. GEM gas detectors in photon counting mode (noise free) were developed in the range 1-30 keV having high efficiency, high time resolution (up to microseconds), energy discrimination in bands and optical flexibility. Discrimination of X-rays, neutrons and gammas has been demonstrated, thanks to the combination of intrinsic gain and discrimination thresholds, at neutron fluxes (10{sup 7} n/s*cm{sup 2}) comparable with the expected ones at the ports of ITER. GEM detectors are also extremely flexible in the design, allowing optimization of the measurements and solutions for shielding or minimization of the effect of background radiation. Two solid state C-MOS imagers working in photon counting mode, one based on Si semiconductor (Medipix-2, range 5-30 keV) and the other one having a CdTe sensor (Pixirad, range 2-100 keV) have been characterized in laboratory. C-MOS imagers have features and performances thoroughly complementary to the GEM detector and, thanks to their higher pixel density, could be used as remote imaging detectors coupled to optics. X-ray polycapillary lenses have been therefore studied in the laboratory both as imaging optical device (full lens) or to define a line of sight (cylinder lens); the preliminary results are encouraging toward the goal of using these lenses to transport X-ray radiation far from the reactor. Reflective or diffractive X-ray optics can be also an option for tomography. A GEM detector has been installed at KSTAR and in the next future hopefully the other approaches will be tested there: once their feasibility will be demonstrated, the issue of the improvement of the radiation tolerance will be faced and the development of algorithms for data analysis as well.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Suppression of Tearing Modes by Radio Frequency Current Condensation [Suppression of Tearing Modes by RF Current Condensation]

Currents driven by radio frequency (rf) waves in the interior of magnetic islands can stabilize deleterious tearing modes in tokamaks. Present analyses of stabilization assume that the local electron acceleration is unaffected by the presence of the island. However, the power deposition and electron acceleration are sensitive to the perturbation of the temperature. The nonlinear feedback on the power deposition in the island increases the temperature perturbation, and can lead to a bifurcation of the solution to the steady-state heat diffusion equation. The combination of the nonlinearly enhanced temperature perturbation with the rf current drive sensitivity to the temperature leads to an rf current condensation effect, which can increase the efficiency of rf current drive stabilization and reduce its sensitivity to radial misalignment of the ray trajectories. As a result, the threshold for the effect is in a regime that has been encountered in experiments, and will likely be encountered in ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

L–H transition threshold studies in helium plasmas at JET

We present a study of the power threshold for L–H transitions (P LH ) in almost pure helium plasmas, obtained in recent experiments at JET with an ITER-like wall (Be wall and W divertor). The most notable new result is that the density at which P LH is minimum, $\bar{n}_{c,min}$, is considerably higher for helium than for deuterium and hydrogen plasmas. We discuss the possible implications for ITER in its pre-fusion operating power phase.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Time-dependent SOLPS-ITER simulations of the tokamak plasma boundary for model predictive control using SINDy *

Abstract Time-dependent SOLPS-ITER simulations have been used to identify reduced models with the sparse identification of nonlinear dynamics (SINDy) method and develop model-predictive control of the boundary plasma state using main ion gas puff actuation. A series of gas actuation sequences are input into SOLPS-ITER to produce a dynamic response in upstream and divertor plasma quantities. The SINDy method is applied to identify reduced linear and nonlinear models for the electron density at the outboard midplane n e , s e p O M P and the electron temperature at the outer divertor T e , s e p d i v . Note that T e , s e p d i v is not necessarily the peak value of T e along the divertor. The identified reduced models are interpretable by construction (i.e. not black box), and have the form of coupled ordinary differential equations. Despite significant noise in T e , s e p d i v , the reduced models can be used to predict the response over a range of actuation levels to a maximum deviation of 0.5% in n e , s e p O M P and 5%–10% in T e , s e p d i v for the cases considered. Model retraining using time history data triggered by a preset error threshold is also demonstrated. A model predictive control strategy for nonlinear models is developed and used to perform feedback control of a SOLPS-ITER simulation to produce a setpoint trajectory in n e , s e p O M P using the integrated plasma simulator framework. The developed techniques are general and can be applied to time-dependent data from other boundary simulations or experimental data. Ongoing work is extending the approach to model identification and control for divertor detachment, which will present transient nonlinear behavior from impurity seeding, including realistic latency and synthetic diagnostic signals derived from the full SOLPS-ITER output.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Isotope effects on L-H threshold and confinement in tokamak plasmas

The dependence of plasma transport and confinement on the main hydrogenic ion isotope mass is of fundamental importance for understanding turbulent transport and, therefore, for accurate extrapolations of confinement from present tokamak experiments, which typically use a single hydrogen isotope, to burning plasmas such as ITER, which will operate in deuterium–tritium mixtures. Knowledge of the dependence of plasma properties and edge transport barrier formation on main ion species is critical in view of the initial, low-activation phase of ITER operations in hydrogen or helium and of its implications on the subsequent operation in deuterium–tritium. The favourable scaling of global energy confinement time with isotope mass, which has been observed in many tokamak experiments, remains largely unexplained theoretically. Moreover, the mass scaling observed in experiments varies depending on the plasma edge conditions. In preparation for upcoming deuterium–tritium experiments in the JET tokamak with the ITER-like Be/W Wall (JET-ILW), a thorough experimental investigation of isotope effects in hydrogen, deuterium and tritium plasmas is being carried out, in order to provide stringent tests of plasma energy, particle and momentum transport models. Recent hydrogen and deuterium isotope experiments in JET-ILW on L-H power threshold, L-mode and H-mode confinement are reviewed and discussed in the context of past and more recent isotope experiments in tokamak plasmas, highlighting common elements as well as contrasting observations that have been reported. The experimental findings are discussed in the context of fundamental aspects of plasma transport models.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Helical core formation and evolution during current ramp-up in the high-field tokamak Alcator C-Mod

We report that large, spontaneous m/n = 1/1 helical cores are predicted in tokamaks with extended regions of low- or reversed-magnetic shear profiles in a region within the q = 1 surface and an onset condition determined by constant (dp/dρ)/B$2\atop{t}$ along the threshold. These 3D modes occurred frequently in Alcator C-Mod during ramp-up when slow current penetration results in a reversed shear q-profile. The onset and early development of a helical core in C-Mod were simulated using a new 3D time-dependent equilibrium reconstruction, based on the ideal MHD equilibrium code VMEC. The reconstruction used the experimental density, temperature, and soft-X-ray fluctuations. The pressure profile can become hollow due to an inverted, hollow electron temperature profile caused by molybdenum radiation in the plasma core during the current ramp-up phase before the onset of sawteeth, which may also occur in ITER with tungsten. Lastly, based on modeling, it is found that a reverse shear q-profile combined with a hollow pressure profile reduces the onset condition threshold, enabling helical core formation from an otherwise axisymmetric equilibrium.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Parameter dependencies of the experimental nitrogen concentration required for detachment on ASDEX Upgrade and JET

While current tokamak experiments are beginning to use real-time feedback control systems to manage the plasma exhaust, future tokamaks still require validation of theoretical models used to predict the threshold impurity concentration required to sufficiently reduce the power and particle fluxes to the divertor. This work exploits new spectroscopic measurements of the divertor nitrogen concentration, c n , in partially detached N 2 -seeded H-mode plasmas on ASDEX Upgrade (AUG) and JET with the ITER-Like Wall (JET-ILW) to test the parameter dependencies of the power flowing to the outer divertor, P div,outer , and the separatrix electron density, n e.sep . A least-squares regression of the AUG measurements demonstrates that the threshold c n required for detachment scales as $c_n∝P^{1.19±0.32}_{div,outer}n^{-2.77±0.36}_{e,sep}$. This scaling of n e,sep is also consistent with the measurements from JET which, at constant P div,outer , show $c_n∝n^{-2.43±0.27}_{e,sep}$. The dependencies of P div,outer and n e.sep is demonstrated over at least a factor of two change in both parameters and indicates a stronger dependence on n e.sep in comparison to the Lengyel model, which could be due to the assumption in this model that the heat flux channel width is independent of density. This first assessment of detachment scaling with impurity seeding highlights the need for further analysis of the systematic uncertainties of the measurement and more consistent scenarios from more tokamaks to investigate the machine size scaling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗