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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Overview of Liquid-Metal PFC R&D at the University of Illinois Urbana-Champaign

The design and implementation of future flowing liquid-lithium plasma-facing components (LLPFCs) will be dependent on several factors. Of course, one of the most important is the need to be able to deal with high heat fluxes incident on the surface of the LLPFCs, but there are also several other important liquid-metal behaviors that have been identified for their critical impact on the feasibility of a LLPFC. One of these is the ability to constantly wet 100% of the plasma-facing component area and the best way to achieve that. Another key point is knowing and understanding the erosion and corrosion of the surfaces subject to a flowing liquid-lithium system and the ability for hydrogen and helium uptake by the system. The Center for Plasma Material Interactions (CPMI) has been tasked with looking at these various issues. The Mock-up Entry module for EAST device was used to investigate wetting and erosion effects and to design a suitable distribution and collection system with a liquid-lithium loop. The vapor shielding effects of lithium on the surface were also modeled and studied. A model coupling CRANE, an open-source global reaction network solver, and Zapdos, a plasma transport solver, is being developed to better understand the dynamics of the vapor cloud. Experiments on the Magnum-PSI at the Dutch Institute for Fundamental Energy Research have been carried out to study the vapor shielding effect and obtain experimental benchmarks to verify the model. Also, initial experiments using the Hybrid Illinois Device for Research and Applications have been performed to understand the pumping effects of lithium on helium. Experiments with a drop of liquid lithium (~100 mg) into a helium plasma have shown the ability of lithium to take out the cold recycling helium gas as well as hydrogen and oxygen impurity gases. The improvement in plasma performance was significant, and further understanding of this effect will have impacts on how future LLPFCs will be designed. Further investigation into the exact mechanism for helium pumping by lithium needs to be performed in the future. Finally, this paper presents a summary of the results obtained at the CPMI.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Fast neural Poincaré maps for toroidal magnetic fields

Poincaré maps for toroidal magnetic fields are routinely employed to study gross confinement properties in devices built to contain hot plasmas. In most practical applications, evaluating a Poincaré map requires numerical integration of a magnetic field line, a process that can be slow and that cannot be easily accelerated using parallel computations. We propose a novel neural network architecture, the HénonNet, and show that it is capable of accurately learning realistic Poincaré maps from observations of a conventional field-line-following algorithm. After training, such learned Poincaré maps evaluate much faster than the field-line integration method. Moreover, the HénonNet architecture exactly reproduces the primary physics constraint imposed on field-line Poincaré maps: flux preservation. Furthermore, this structure-preserving property is the consequence of each layer in a HénonNet being a symplectic map. We demonstrate empirically that a HénonNet can learn to mock the confinement properties of a large magnetic island by using coiled hyperbolic invariant manifolds to produce a sticky chaotic region at the desired island location. This suggests a novel approach to designing magnetic fields with good confinement properties that may be more flexible than ensuring confinement using KAM tori.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Edge localized mode characteristics and divertor heat flux during stationary and transient phase for CFETR hybrid scenario

The study of edge localized mode (ELM) behavior, stationary heat flux and transient heat flux during the ELM crash phase is performed for a China Fusion Engineering Test Reactor (CFETR) 1 GW hybrid mode operation scenario ($R$ = 7.2 m, ${B_{\text{T}}}$ = 6.5 T, ${I_{\text{p}}}$ = 13.78 MA). Modeling and simulation start with a scenario obtained by multi-code integrated modeling on the one modeling framework for integrated tasks framework. Linear stability and nonlinear simulations of ELM dynamics are carried out using the BOUT++ six-field reduced magnetohydrodynamic module, which show a much smaller ELM energy loss (Δ ELM ~ 0.13%) compared to that of a Type-I ELM. Parametric analysis of the weak linear growth rate and small ELM energy loss characteristics shed light on physics corresponding to a grassy ELM regime for CFETR 1 GW hybrid scenario. The transient heat flux on the divertor target during this small ELM phase is investigated using BOUT++. Here we found that upstream radial transport in the scrape-off-layer (SOL) induced by small ELMs is weak, which keeps it in the drift-dominated region. However, the heat flux width is still broadened to ${\lambda _{\text{q}}}$ = 4.64 mm by the increase of separatrix temperature during ELM nonlinear evolution. The impact of transient peak heat load and ELM energy fluence on tungsten melting and net erosion rate of divertor target is evaluated for the first-time using physics-based transport that connects the pedestal with the SOL. Energy fluence caused by a single ELM pulse is below the tungsten melting limit, while tungsten erosion would exceed the material requirements. We conclude that external mitigation methods, such as divertor detachment and advanced divertor geometry are likely needed for the steady state operation of CFETR.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Single-stage stellarator optimization: combining coils with fixed boundary equilibria

We introduce a novel approach for the simultaneous optimization of plasma physics and coil engineering objectives using fixed-boundary equilibria that is computationally efficient and applicable to a broad range of vacuum and finite plasma pressure scenarios. Our approach treats the plasma boundary and coil shapes as independently optimized variables, penalizing the mismatch between the two using a quadratic flux term in the objective function. Four use cases are presented to demonstrate the effectiveness of the approach, including simple and complex stellarator geometries. As shown here, this method outperforms previous two-stage approaches, achieving smaller plasma objective function values when coils are taken into account.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

High-performance plasmas after pellet injections in Wendelstein 7-X

A significant improvement of plasma parameters in the optimized stellarator W7-X is found after injections of frozen hydrogen pellets. The ion temperature in the post-pellet phase exceeds 3 keV with 5 MW of electron heating and the global energy confinement time surpasses the empirical ISS04-scaling. The plasma parameters realized in such experiments are significantly above those in comparable gas-fuelled discharges. Here, we present details of these pellet experiments and discuss the main plasma properties during the enhanced confinement phases. Local power balance is applied to show that the heat transport in post-pellet phases is close to the neoclassical level for the ion channel and is about a factor of two above that level for the combined losses. In comparable gas-fuelled discharges, the heat transport is by about ten times larger than the neoclassical level, and thus is largely anomalous. It is further observed that the improvement in the transport is related to the peaked density profiles that lead to a stabilization of the ion-scale turbulence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Overview of the results from divertor experiments with attached and detached plasmas at Wendelstein 7-X and their implications for steady-state operation

Abstract Wendelstein 7-X (W7-X), the largest advanced stellarator, is built to demonstrate high power, high performance quasi-continuous operation. Therefore, in the recent campaign, experiments were performed to prepare for long pulse operation, addressing three critical issues: the development of stable detachment, control of the heat and particle exhaust, and the impact of leading edges on plasma performance. The heat and particle exhaust in W7-X is realized with the help of an island divertor, which utilizes large magnetic islands at the plasma boundary. This concept shows very efficient heat flux spreading and favourable scaling with input power. Experiments performed to overload leading edges showed that the island divertor yields good impurity screening. A highlight of the recent campaign was a robust detachment scenario, which allowed reducing power loads even by a factor of ten. At the same time, neutral pressures at the pumping gap entrance yielded the particle removal rate close to the values required for stable density control in steady-state operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

The constraint of plasma power balance on runaway avoidance

Abstract In a post-thermal-quench plasma, mitigated or unmitigated, the plasma power balance is mostly between collisional or Ohmic heating and plasma radiative cooling. In a plasma of atomic mixture { n α } with α labeling the atomic species, the power balance sets the plasma temperature, ion charge state distribution { n α i } with i the charge number, and through the electron temperature T e and ion charge state distribution { n α i } , the parallel electric field E ∥ . Since the threshold electric field for runaway avalanche growth E a v is also set by the atomic mixture, ion charge state distribution and its derived quantity, the electron density n e , the plasma power balance between Ohmic heating and radiative cooling imposes a stringent constraint on the plasma regime for avoiding and minimizing runaways when a fusion-grade tokamak plasma is rapidly terminated.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Enhanced plasma performance in C-2W advanced beam-driven field-reversed configuration experiments

TAE Technologies’ fifth-generation fusion device, C-2W (also called ‘Norman’), is the world’s largest compact-toroid device and has made significant progress in field-reversed configuration (FRC) plasma performance. C-2W produces record breaking, macroscopically stable, high-temperature advanced beam-driven FRC plasmas, dominated by injected fast particles and sustained in steady state, which is primarily limited by neutral-beam (NB) pulse duration. The NB power supply system has recently been upgraded to extend the pulse length from 30 ms to 40 ms, which allows for a longer plasma lifetime and thus better characterization and further enhancement of FRC performance. An active plasma control system is routinely used in C-2W to produce consistent FRC performance as well as for reliable machine operations using magnet coils, edge-biasing electrodes, gas injection and tunable-energy NBs. Google’s machine learning framework for experimental optimization has also been routinely used to enhance plasma performance. D

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Corrigendum: Simulations of stand-off runaway electron beam termination by tungsten particulates for tokamak disruption mitigation (2024 Nucl. Fusion 64 056019)

In the original article, we inadvertently omitted a reference which previously introduced the concept of tungsten injection for 'Disruption Mitigation in Tokamak Reactor via Reducing the Seed Electrons of Avalanche.' In that reference, the authors have proposed injection of a tungsten cylinder of 80 mm by 8 mm dimensions, by rail guns, to absorb the runaway seed population. This can be contrasted with another interesting idea of using small tungsten pellets coated with a low Z material for depleting the runaway seeds as an option for upgrade of the ITER Disruption Mitigation System. Interested readers are referred to those two papers on the tungsten injection for runaway seed removal/reduction, and the related issue of not shortening the current quench excessively. The application focus of was not on reducing the runaway seeds in the initial Ohmic-to-runaway current conversion phase, but on safe termination of a fully formed and likely decaying runaway beam that is about to scrape off against the first wall, which can be accelerated by vertical displacement events. This scheme thus serves as a last line of defense against potential wall damage. The choice here is locally released tungsten particulates, similar to the previous idea of using tungsten particulates as a dust shield for the divertor. The most significant finding of is that strong pitch angle scattering, in addition to energy attenuation and absorption, of high-energy runaways by tungsten particulates, provides a transport mechanism by which runaways damage of the first wall can be mitigated by reducing and spreading the runaway wall load. We regret the omission in the original article, and thank the authors of the cited article for bringing this matter to our attention. We also find the possibility of using the injected tungsten rod from to terminate the runaways, as opposed to the cloud of tungsten particulates in intriguing. In that context, it is of interest to mention another possibility of a retractable tungsten metal arm that can be swung out for deployment in runaway termination, which would remove the need for post-mitigation recovery of tungsten debris in injection schemes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Electron transport barrier and high confinement in configurations with internal islands close to the plasma edge of W7-X

The low magnetic shear in the Wendelstein 7-X (W7-X) stellarator makes it feasible to shape the separatrix by the large islands constituting an island-divertor, and this can be exploited to access various magnetic configurations, including samples of different internal island sizes and locations. To investigate the configuration effects on the plasma confinement, a configuration scan was performed by changing the coil currents to vary the rotational transform between values 5/4 and 5/6 at the plasma boundary with different power levels (2, 4, 6 MW) of electron cyclotron resonance heating (ECRH) at a maximum plasma density of 8 × 10 19 m -2 . neutral beam injection (NBI) heating was also applied during some configurations of the scan to create a density ramp and access high densities beyond the X2 ECRH cutoff. For the magnetic configurations, where the 5/5 and 5/6 island chains were moved closer to separatrix but remaining inside the last closed flux surface, the electron cyclotron emission shows that an electron temperature, T e , pedestal develops already during ECRH heated plasma buildup phase indicating a transport barrier, and the barrier sustains irrespective of changed plasma heating conditions such as NBI in the later part of discharge. The transport barrier is broken by subsequent fast crashes, observed with multiple plasma diagnostics with characteristics such as tokamak edge localized modes, and the corresponding crash amplitude and frequency vary with plasma pressure. The impact of the transport barrier on plasma confinement can be seen through the increased core T e profile, which could be responsible for the overall increase in the stored diamagnetic energy by approximately 10% for these configurations. After the plasma heating is terminated, a backwards transition to a degraded confinement state is also observed. These observations indicate a configuration triggered high confinement mode in low shear W7-X. This work focuses on the occurrence of this transport barrier for different magnetic configurations and its relation to internal magnetic islands.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Marginally stable current sheets in collisionless magnetic reconnection

Noncollisional current sheets that form during the nonlinear development of spontaneous magnetic reconnection are characterized by a small thickness, of the order of the electron skin depth. They can become unstable to the formation of plasmoids, which allows the magnetic reconnection process to reach high reconnection rates. In this work, we investigate the marginal stability conditions for the development of plasmoids when the forming current sheet is purely collisionless and in the presence of a strong guide field. We analyze the geometry that characterizes the reconnecting current sheet, and what promotes its elongation. Once the reconnecting current sheet is formed, we identify the regimes for which it is plasmoid unstable. Finally, our study shows that plasmoids can be obtained, in this context, from current sheets with an aspect ratio much smaller than in the collisional regime, and that the plasma flow channel of the marginally stable current layers maintains an inverse aspect ratio of 0.1.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Runaway electron current reconstitution after a nonaxisymmetric magnetohydrodynamic flush

Benign termination of mega-ampere (MA) level runaway current has been convincingly demonstrated in recent JET and DIII-D experiments, establishing it as a leading candidate for runaway mitigation on ITER. This comes in the form of a runaway flush by parallel streaming loss along stochastic magnetic field lines formed by global magnetohydrodynamic instabilities, which are found to correlate with a low-Z injection that purges the high-Z impurities from a post-thermal-quench plasma. Here, in this study, we show the competing physics that govern the postflush reconstitution of the runaway current in an ITER-like reactor where significantly higher current is expected. The trapped “runaways” are found to dominate the seeding for runaway reconstitution, and the incomplete purge of high-Z impurities helps drain the seed but produces a more efficient avalanche, two of which compete to produce a 2–3 MA step in current drop before runaway reconstitution of the plasma current.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Testing the Activation Analysis for Fusion in OpenMC

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It can perform fission simulations such as fixed-source, k-eigenvalue, and subcritical multiplication calculations on models built using either a constructive solid geometry or CAD representation. To explore the use of OpenMC for fusion activation analysis, a detailed model of the Fusion Neutronics Science Facility (FNSF) was first developed for comparisons against an existing SERPENT model. A 90-degree model of FNSF in Standard-Triangle-Language (STL) CAD format was converted to Constructive Solid Geometry (CSG) using each code's built-in functions, and the geometries were validated by ensuring no cells overlapped and no particles were lost during simulations. The neutron fluxes were calculated and compared for multiple components close to the plasma. The results show differences mostly below 1% in fluxes and averaged 8% for activity and decay heat. Here, the work described in this study tests the CAD-based geometry using the DagMC toolkit in OpenMC and compares the activation analysis of OpenMC to SERPENT code.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Cooling flow regime of a plasma thermal quench

A large class of Laboratory, Space, and Astrophysical plasmas is nearly collisionless. When a localized energy or particle sink, for example, in the form of a radiative cooling spot or a black hole, is introduced into such a plasma, it can trigger a plasma thermal collapse, also known as a thermal quench in tokamak fusion. Here we show that the electron thermal conduction in such a nearly collisionless plasma follows the convective energy transport scaling in itself or in its spatial gradient, due to the constraint of ambipolar transport. As a result, a robust cooling flow aggregates mass toward the cooling spot and the thermal collapse of the surrounding plasma takes the form of four propagating fronts that originate from the radiative cooling spot, along the magnetic field line in a magnetized plasma. The slowest one, which is responsible for deep cooling, is a shock front.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Modeling a Hypothetical Fusion Blanket Design Using MELCOR-TMAP

MELCOR (not an acronym) is a nuclear safety code developed by Sandia National Laboratories for the Nuclear Regulatory Commission. Idaho National Laboratory collaborates with Sandia to maintain and develop a version of MELCOR specifically for fusion device applications. Most recently, the tritium migration analysis program (TMAP) has been incorporated into MELCOR for fusion, in a version called MELCOR-TMAP. We show the results applying MELCOR-TMAP to model steady-state inventory and tritium retention in a liquid immersion blanket design for a hypothetical fusion device under steady-state operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗