Early Nuclear Fusion Cross-Section Advances 1934–1952 and Comparison to Today’s ENDF Data
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70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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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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Some of the History Surrounding the Oliphant et al. Discovery of dd Fusion and an Inference of the d(d,p)t Cross Section from This 1934 Paper
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Introduction to Special Issue on the Early History of Nuclear Fusion
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The Untold Story of Building the First Megaton Thermonuclear Fusion Device: The Simple Element and IVY Mike
This report is a research and development engineer’s perspective on the fascinating story of the world’s first megaton-class thermonuclear device, IVY Mike (10.4 Mt). Few modern scientific endeavors have matched the complexity and breadth of scientific achievement in such a short amount of time as IVY Mike. This paper will take a look at the design, engineering, fielding, and execution of the world’s first megaton-class physics experiment by Los Alamos Scientific Laboratory and the birth of industrial liquid hydrogen use spawned by the Cold War effort. Although others have written on aspects of this technical history, they have not benefited from access to the original classified documents used by the present author. The present paper must necessarily omit some technical details that remain classified but represents the most comprehensive summary of the engineering and fabrication of the IVY Mike device in the open literature.
45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
Staged cooling of a fusion-grade plasma in a tokamak thermal quench
Abstract In tokamak disruptions where the magnetic connection length becomes comparable to or even shorter than the plasma mean-free-path, parallel transport can dominate the energy loss and the thermal quench of the core plasma goes through four phases (stages) that have distinct temperature ranges and durations. The main temperature drop occurs while the core plasma remains nearly collisionless, with the parallel electron temperature T e ∥ dropping in time t as T e ∥ ∝ t − 2 and a cooling time that scales with the ion sound wave transit time over the length of the open magnetic field line. These surprising physics scalings are the result of effective suppression of parallel electron thermal conduction in an otherwise bounded, quasineutral, and collisionless plasma, which is different from what are known to date on electron thermal conduction along the magnetic field in a nearly collisionless and quasineural plasma.
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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.
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Measurements of the temperature and velocity of the dense fuel layer in inertial confinement fusion experiments
Abstract not provided.
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First Evidence of Local E × B Drift in the Divertor Influencing the Structure and Stability of Confined Plasma near the Edge of Fusion Devices
The structure of the edge plasma in a magnetic confinement system has a strong impact on the overall plasma performance. Here we uncover for the first time a magnetic-field-direction dependent density shelf, i.e., local flattening of density radial profile near the magnetic separatrix, in high confinement plasmas with low edge collisionality in the DIII-D tokamak. The density shelf is correlated with a doubly peaked density profile near the divertor target plate, which tends to occur for operation with the ion BX∇B drift direction away from the X-point, as currently employed for DIII-D advanced tokamak scenarios. This double-peaked divertor plasma profile is connected via the EXB drifts, arising from a strong radial electric field induced by the radial electron temperature gradient near the divertor target. The drifts lead to the reversal of the poloidal flow above the divertor target, resulting in the formation of the density shelf. The edge density shelf can be also enhanced at higher heating power, preventing large, periodic bursts of the plasma, i.e. Edge-localized Modes, in the edge region, consistent with ideal magnetohydrodynamics calculations.
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Detachment in Fusion Plasmas with Symmetry Breaking Magnetic Perturbation Fields
Power exhaust from the bulk plasma is significantly altered by symmetry breaking magnetic perturbation fields, because these create direct connections (perturbed field lines) from the confined high temperature plasma to solid surfaces. Additionally, the same amount of power is distributed among those new exhaust channels as for a symmetric magnetic configuration, which reduces the local upstream heat flux flowing down the perturbed field lines, thereby making access to detachment easier (i.e., at lower upstream density) for the divertor plasma near the location corresponding to the symmetric magnetic separatrix. However, the divertor plasma regions with connection to the bulk plasma are extended nonaxisymmetrically further outside, where significant heat loads occur, unlike in the symmetric configuration. The temperature remains high at those locations, which reduces the divertor plasma dissipation capacity, making the mitigation of heat loads more difficult to achieve.
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Experimental Validation of a Kinetic Ballooning Mode in High-Performance High-Bootstrap Current Fraction Fusion Plasmas
We report the observation of a set of coherent high frequency electromagnetic fluctuations that leads to a turbulence induced self-regulating phenomenon in the DIII-D high bootstrap current fraction plasma. The fluctuations have frequency of 130~220kHz, the poloidal wave length and phase velocity are 16~30 m -1 and ~30 km/s, respectively in the outboard midplane with the estimated toroidal mode number n~5- 9. The fluctuations are located in the internal transport barrier (ITB) region at large radius and are experimentally validated to be kinetic ballooning modes (KBM). Furthermore, quasilinear estimation predicts the KBM to be able to drive experimental particle flux and non-negligible thermal flux, suggesting its significant role in regulating the ITB saturation.
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Broadening of the Divertor Heat Flux Profile in High Confinement Tokamak Fusion Plasmas with Edge Pedestals Limited by Turbulence in DIII-D
Multimachine empirical scaling predicts an extremely narrow heat exhaust layer in future high magnetic field tokamaks, producing high power densities that require mitigation. Here, in the experiments presented, the width of this exhaust layer is nearly doubled using actuators to increase turbulent transport in the plasma edge. This is achieved in low collisionality, high confinement edge pedestals with their gradients limited by turbulent transport instead of large-scale, coherent instabilities. The exhaust heat flux profile width and divertor leg diffusive spreading both double as a high frequency band of turbulent fluctuations propagating in the electron diamagnetic direction doubles in amplitude. The results are quantitatively reproduced in electromagnetic XGC particle-in-cell simulations which show the heat flux carried by electrons emerges to broaden the heat flux profile, directly supported by Langmuir probe measurements.
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Persistent Hot-Spot Mix in Cryogenic Direct-Drive Fusion Experiments
We show that an x-ray emission signature associated with acceleration phase mass injection correlates with poor experimental hot-spot convergence and a reduced neutron production relative to expectations. It is shown that with increased target mass as well as with higher-design adiabats, this signature is reduced, whereas with increased debris on the target the signature is increased. Here, we estimate that the vapor region in typical best designs may have up to 2x the assumed hydrogen mass at the start of deceleration.
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Benchmarking solid-to-plasma transition modeling for inertial confinement fusion laser-imprint with a pump-probe experiment
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Bus Design for the Poloidal Field Coils of the NSTX-Upgrade Fusion Device
The National Spherical Torus eXperiment (NSTX) has undergone a major upgrade to NSTX-U at Princeton Plasma Physics Laboratory (PPPL). NSTX upgrade (NSTX-U) will double the toroidal field, plasma current, and neutral beam injection heating power, as well as significantly increase the pulse duration. NSTX-U uses three poloidal field (PF) coils at the vessel top and three at the bottom near the divertor areas to control the local plasma shape there. These shaping coils operate at a maximum terminal voltage of 2 kV, corresponding to a maximum current about 20 kA. To supply the power to the PF coils, hard copper bus bars are typically used near the vacuum vessel, while water cooled flexible cables are used away from the vessel. The detailed design of the hard copper bus bars will be covered in this article. During operation, these hard bus bars are subject to high electromagnetic (EM) forces, thermal displacement loads, as well as plasma halo loads due to disruption. The EM, thermal, and structural analysis were performed, and the results revealed that, with the worst loads combined, the new design meets the NSTX-U thermal, structural, and fatigue cycle requirements. Finally, the manufacturing and installation process of the bus work will be discussed at the end of this article.
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Careers in Fusion and High Energy Density science at Los Alamos National Laboratory [Slides]
Abstract not provided.
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Machine Learning for Real-time Fusion Plasma Behavior Prediction and Manipulation
This project set out with an ambitious goal: to develop and apply machine learning-based methods of discovering new controllers and operating regimes for achieving better performing plasmas in tokamaks. As summarized in the dozens of papers below, the project was a huge success. It developed several machine learning components and integrated them into a single ML system that was added to the DIII-D PCS. That functionality was demonstrated over a series of experiments on DIII-D. A summary of these results is given here with the details in the published papers below.
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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.
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Gyrokinetic Particle Simulations of Core Turbulence in Fusion Plasmas
In this project, we have further developed GTC gyrokinetic particle simulations to study plasma confinement properties in the stellarators and tokamaks with 3D magnetic fields, where both neoclassical and turbulent transport are self-consistently treated on an equal footing.
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