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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Estimates of global recycling coefficients for LTX- β discharges

We report the first observation of global recycling coefficient R near 0.5 in the Lithium Tokamak eXperiment-β (LTX-β), significantly below the minimum R previously reported in other devices. In a series of experiments with varied Li wall conditioning, estimates of the recycling coefficient have been made using a Lyman-α array and DEGAS2 modeling. A progressive reduction in Lyman-α emission with increased lithium and an increase in edge electron temperature are observed. It is also observed that with increasing Li coating thickness, the effective particle confinement time τp* is reduced and approaches TRANSP calculated energy confinement time (τE), with τp* near τE,TRANSP for the lowest recycling coefficients. Edge temperatures approaching core plasma temperatures, first reported in LTX, can now be directly connected to estimates of the recycling coefficient and qualitatively agree with previous UEDGE simulations. The particle flux to the limiting surfaces appears to be significantly reduced in comparison with fluid scrape-off layer (SOL) models, indicating that a large fraction of the SOL ions are mirror trapped. SOL collisionality drops more than an order of magnitude below the banana regime boundary, indicating the importance of kinetic effects. Full-f 1x2v gyrokinetic simulations of SOL field lines with the GKEYLL code indicate that the fraction of ions trapped along field lines increases as collisionality drops, as a result of increased lithium evaporation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

General kinetic ion-induced electron emission model for metallic walls applied to biased Z-pinch electrodes

A kinetic ion-induced electron emission (IIEE) model for general applications is developed to obtain the emitted electron energy spectrum for a distribution of ion impacts on a metallic surface. We assume an ionization cascade mechanism and use empirical models for the ion and electron stopping powers. The emission spectrum and the secondary electron yield (SEY) are validated for a variety of materials. The IIEE model is used to study the effect of IIEE on the plasma-material interactions of Z-pinch electrodes. Un-magnetized Boltzmann-Poisson simulations are performed for a Z-pinch plasma doubly bounded by two biased copper electrodes with and without IIEE at bias potentials from 0 to 9 kV. At the anode, the SEY decreases from 0 to 1 kV, but then increases at higher bias potentials. At the cathode, the SEY is much larger due to higher energy ion bombardment and grows with bias potential. As the bias potential increases, the emitted cathode electrons are accelerated to higher energies into the domain, collisionally heating the plasma. Above 1 kV, the heating is strong enough to increase the plasma potential. Despite SEY greater than 1, only a classical sheath forms as opposed to a space-charge limited or inverse sheath due to the emitted electron flux not reaching the space charge current saturation limits. Furthermore, the current in the emissionless cases saturates to a value lower than experiment. With IIEE, the current does not saturate and continues to increase with the 4 kV case, matching most closely with the experiment.

Carbon based materials↗
DOE OSTI2025

Molybdenum density profiles in Alcator C-Mod plasmas

Molybdenum (Z = 42) density spatial profiles have been obtained for Alcator C-Mod L-, I-, and H-mode plasmas from x-ray observations of the neon-like Mo 32+ 4D transition. Profile shapes range from flat to modestly peaked in the core. The observed peaking factors for molybdenum are in good agreement with those calculated by TGLF across all confinement regimes, and there is little dependence seen with core density gradient, sawtooth amplitude, or rotation velocity. Predictions for tungsten peaking in SPARC are presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

ORBIT simulations of fast ion power loads on the wall of the Divertor Tokamak Test

Abstract Neutral beam injection in tokamaks produces a population of fast ions, which interact with 3D magnetic fields in a variety of ways, often resulting in energetic particle losses in very short times to the wall. Careful design of neutral beams and active control of error fields helps to keep these losses to a minimum. Nevertheless, past experience in tokamaks in the 1980s and detailed simulations for future machines, such as International Thermonuclear Experimental Reactor, suggest that very localized fast ion losses (‘hot spots’) can be present, even if the overall losses are low. In this paper, we discuss this issue in the Divertor Tokamak Test (DTT) project, and we show that in the standard single-null full-power scenario of the DTT, fast ions produce two hot spots, corresponding approximately to the beam injection and exit toroidal angles: the former being mainly due to prompt losses/passing particles, while the latter is due to trapped ions. However, the maximum power load in these spots is of the order ∼ 63 kW m −2 , below the tolerance of plasma-facing components of the machine.

Physics↗
DOE OSTI2023

Energetics of silicon in the bulk and near surfaces of tungsten: a first-principles study

Abstract Siliconization of the tokamak walls is a candidate method to improve plasma confinement in fusion tokamaks containing tungsten plasma facing components (W PFCs). To understand the interactions of silicon (Si) with W, the Si behavior in bulk W, and near three low-index W surfaces ((100), (110) and (111)) has been investigated using first-principles density functional theory. In bulk W, Si interstitial atoms have a low solution ability and high mobility, and Si atoms can be strongly trapped by W vacancies. The interaction between two Si adatoms is responsible for the stability of adatom superstructures on W surfaces, consistent with previous experimental observation (Tsong and Casanova 1981 Phys. Rev. Lett. 47 113). Although the coverage dependence of Si adsorption and diffusion energetics on surfaces is related to surface orientation, the W(110) surface has lower Si adsorption affinity and higher Si diffusivity than either the W(111) or W(100) surfaces. The most stable Si adatom superstructure on W surfaces is: square c(2 × 2) pattern on W(100) covered with 0.5 ML Si; rectangular c(4 × 2) pattern on W(110) with 0.25 ML Si; and rhombus p(1 × 1) pattern on W(111) with 1 ML Si. The coverage dependence of Si mobility on/toward W surfaces is generally related to the stability of the Si superstructures as a function of coverage on each surface. Interestingly, Si adatoms prefer to transport below the surface and into W subsurface by an exchange mechanism with W atoms, indicating the likelihood of epitaxial growth of W silicide layers on W surfaces during the operation of W PFCs.

Yang, L. (ORCID:0000000322166071)↗
DOE OSTI2024

A semi-automated algorithm for designing stellarator divertor and limiter plates and application to HSX

We present a semi-automated algorithm for designing three-dimensional divertor or limiter plates targeting low heat loads. The algorithm designs the plates in two stages: firstly, the parallel heat flux distribution is caught on vertically-inclined plates at one or several toroidal locations. Secondly, the power per unit area is reduced by stretching, tilting and bending the plates toroidally. Heat transport is modelled using the EMC3-Lite code, which uses an anisotropic diffusion model. We apply this scheme to HSX, a medium-sized stellarator located at the University of Wisconsin–Madison. Starting from the current machine with an extended vessel wall, we construct plates which are able to effectively catch and spread the heat for three different magnetic configurations. The scheme has a computational cost in the order of tens of CPU-minutes, making it a powerful tool for semi-automated plasma-facing component design in three-dimensional environments.

anisotropic diffusion↗
DOE OSTI2026

Simulation of Divertor Performance in ST40 Under Dynamic Double-Null Plasmas

A power fraction model was implemented for the simultaneous prediction of 3-D surface temperature evolution at all four divertor targets in near-double-null (DN) tokamak configurations, which is especially important for compact high-field devices that may not have the ability to dissipate large amounts of power on the high-field side. Evaluating the power-sharing between the four divertor strike points in a disconnected DN configuration is important for understanding the overall power balance, as well as for optimizing the power exhaust performance and prolonging the survivability of the plasma facing components (PFCs). This power-sharing is typically evaluated in terms of the separation between the primary and secondary separatrices at the outboard midplane, $\textit {dR}_{\text {sep}}$. The Heat flux Engineering Analysis Toolkit (HEAT) is coupled with Brunner’s power fraction model to simulate the deposited heat flux and resultant temperature change on 3-D divertor targets in a dynamic DN (DDN) pulse operation in ST40, a high-field spherical tokamak. The simulation results showed that with DDN operation, the operation time has significantly increased compared with single-null geometry configurations.

ST40↗
DOE OSTI2022

Initial Results from Boron Powder Injection Experiments in WEST Lower Single Null L-mode Plasmas

Using a recently installed impurity powder dropper (IPD), boron powder (< 150 μm) was injected into lower single null (LSN) L-mode discharges in WEST. IPDs possibly enable real-time wall conditioning of the plasma-facing components and may help to facilitate H-mode access in the full-tungsten environment of WEST. The discharges in this experiment featured Ip = 0.5 MA, BT = 3.7 T, q95 = 4.3, tpulse = 12–30 s, ne,0 ~ 4×1019 m-2, and PLHCD ~ 4.5 MW. Estimates of the deuterium and impurity particle fluxes, derived from a combination of visible spectroscopy measurements and their corresponding S/XB coefficients, showed decreases of ~ 50% in O+, N+, and C+ populations during powder injection and a moderate reduction of these low-Z impurities (~ 50%) and W (~ 10%) in the discharges that followed powder injection. Along with the improved wall conditions, WEST discharges with B powder injection observed improved confinement, as the stored energy WMHD, neutron rate, and electron temperature Te increased significantly (10–25% for WMHD and 60–200% for the neutron rate) at constant input power. These increases in confinement scale up with the powder drop rate and are likely due to the suppression of ion temperature gradient (ITG) turbulence from changes in Zeff and/or modifications to the electron density profile.

Boronization↗
DOE OSTI2025

High Strength and Fracture Resistance of Reduced-Activity W-Ta-Ti-V-Zr High-Entropy Alloy for Fusion Energy Applications

Refractory high-entropy alloys (HEAs) are promising candidates for next-generation nuclear applications, particularly fusion reactors, due to their excellent high-temperature mechanical properties and irradiation resistance. Here, the microstructure and mechanical behavior were investigated for an equimolar WTaTiVZr HEA, designed from a palette of low-activation elements. The as-cast alloy exhibited a dendritic microstructure composed of W-Ta rich dendrites and Zr-Ti-V rich inter-dendritic regions, both possessing a body-centered cubic (BCC) crystal structure. Room temperature bulk compression tests showed ultra-high strength of around 1.6 GPa and plastic strain ~6%, with fracture surfaces showing cleavage facets. The alloy also demonstrated excellent high-temperature strength of ~650 MPa at 500 °C. Scratch-based fracture toughness was ~38 MPa√m for the as-cast WTaTiVZr HEA compared to ~25 MPa√m for commercially used pure tungsten. This higher value of fracture toughness indicates superior damage tolerance relative to commercially used pure tungsten. These results highlight the alloy’s potential as a low-activation structural material for high-temperature plasma-facing components (PFCs) in fusion reactors.

Physics↗
DOE OSTI2024

THERMAL STRUCTURAL AND SHOCK EVENT EVALUATIONS OF THE FUELING PELLET INJECTION SYSTEM FOR ITER

The US is among seven partner nations in a collaborative effort to design, build, and demonstrate fusion’s ability to be a large-scale carbon free energy source. Each country has its own Domestic Agencies (DA) that contribute directly to the ITER project. US ITER, which is a DOE Office of Science project managed by Oak Ridge National Laboratory, is developing world class engineering solutions to the design, construction, and assembly of the burning plasma experiment that can demonstrate the scientific and technological feasibility of fusion. US ITER’s scope includes completing the preliminary and final design, qualifying materials and processes for manufacture and testing, executing manufacturing, and delivering the Fueling Pellet Injection System (FPIS) to the ITER site for assembly. The US-ITER FPIS system is designed to inject cryogenically frozen pellets of deuterium-tritium (D-T) into the plasma. The FPIS has two main functions: 1. Provide a steady supply of deuterium and tritium fuel, 2. Mitigating the impact of edge localized modes on the plasma facing components. The FPIS will have the capability to reside in three port cells within a tritium second barrier containment cask. Each pellet cask contains three flight tubes linking with penetrations on the torus cryopump housing and vacuum vessel (VV): two for the magnetic high field side (HFS) pellet injection and one for the magnetic low field side (LFS) pellet injection. This paper presents results of the thermal-structural analyses of the FPIS flight tube structural components when subjected to various loads such as electromagnetic (EM) loads, nuclear heating, seismic, and operational, and dynamic shock events. The resulting temperatures and stresses under combined conditions have been found to satisfy the design criteria to ensure safe and reliable operation of the FPIS flight tubes within the vacuum vessel.

Martinez, Oscar↗
DOE OSTI2021

Adaptive complex nanocomposite alloys for burning plasma-material interface tunability (Final Report)

The focus of this program is discovery and development of novel self-healing and adaptive materials for the PMI (plasma-material interface) envisioned for future plasma-burning extreme environments in thermonuclear fusion reactors that can provide enhanced radiation-tolerance or resistance. This program was conceived and stems from the PI’s DOE Early Career Award work on harnessing nanotechnology and mesoscale materials design in refractory metals to address gaps in PMI research. These gaps pertain to the lack of understanding of multi-scale interactions at the plasma-material interface and the development of novel material interfaces that can be designed to adapt to extreme fusion reactor conditions. The final report is separated into two primary sections: The first section consists of the program’s first two years (FY16-FY17) performance period. The second section consists of FY18 to FY19 period, where the FY19 period was a supplemental addition to support the completion of a PhD thesis and continuation of another. This work consists of an excellent team consisting of a co-PI, one postdoctoral researcher, two graduate students, two undergraduate students and collaborators both domestic and abroad.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Thermal-Fluid and Thermal-Structural Response of the T-Tube Modular Divertor to Spatiotemporally Varying Heat Loads

Tungsten (W) is the leading candidate for divertor target plates because of its high melting point (>3000°C), thermal conductivity, and ultimate tensile stress. While W and its alloys are the only solid materials that can survive the high heat fluxes incident on the divertor, W’s low-ductility high ductile-to-brittle transition temperature of ~600°C and relatively low recrystallization temperature (RT) of ~1300°C pose structural (among other) challenges. The objective of this work is to estimate the thermal-fluid and thermal-structural performance of the helium (He)-cooled T-tube divertor, which was originally developed by the Advanced Reactor Innovation and Evaluation Study (ARIES) using numerical simulations. Here, predictions of temperature distributions across the plasma-facing structural component and surface pressures from computational fluid dynamics simulations are used to determine stress distributions using commercial structural finite element modeling software over a range of fusion-relevant conditions. The maximum allowable incident heat fluxes are determined based on the temperature limits imposed by the ITER elastic Structural Design Criteria for In-vessel Components (SDC-IC) and the maximum RT over a range of He mass flow rates and presented in the form of performance design charts. Our recent work found that thermal- structural criteria accounting for the low ductility of W in a finger-type modular divertor constrain the maximum incident heat fluxes to values well below the ITER specifications, and those based on considering only the RT demonstrate that integrated thermal-fluid and elastic structural performance evaluation are required for accurate assessment of divertor performance. This novel analysis of the T-tube considers how nonuniform and transient incident heat fluxes affect its thermal-fluid and thermal-structural performance, as well as the effect of volumetric heating, which can be as great as 27% of the power incident on the divertor surface. The W tile of the T-tube, with its relatively large plasma-facing area of ~15 cm 2 , will likely experience significant spatial variations in incident heat flux. This work therefore assesses whether steady-state incident heat flux profiles with a peak of 10 MW/m 2 and maximum heat flux gradients of 200 MW/m 2 per m exceed the structural limits imposed by the ITER elastic SDC-IC and the maximum RT over a range of fusion-relevant conditions. The effect of transient heat fluxes typical of plasma detachment and reattachment from the target plate due, for example, to gas injection are also evaluated

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

The Material Plasma Exposure eXperiment: Mission and conceptual design

Mastering Plasma Material Interactions (PMI) is key for obtaining a high performance, high duty-cycle and safe operating fusion reactor. Numerous gaps exist in PMI which have to be addressed before a reactor can be built. In particular the lack of data at high ion fluence, fusion reactor divertor relevant plasma conditions and neutron displacement damage requires new experimental devices to be able to develop plasma facing materials and components. This has been recognized in the community and the U.S. fusion program is addressing this need with a new linear plasma device—the Material Plasma Exposure eXperiment (MPEX). MPEX will be a superconducting linear plasma device with magnetic fields of up to 2.5 T. The plasma source is a high-power helicon source (200 kW, 13.56 MHz). The electrons will be heated via Electron Bernstein Waves with microwaves using multiple 70 GHz gyrotrons (up to 600 kW in total). Ions will be heated via ion cyclotron heating in the so-called “magnetic beach heating” scheme in the frequency range of 6–9 MHz (up to 400 kW in total). An overview of the conceptual design and the project/design requirements is given in this paper.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Conference Report on the 8th International Symposium on Liquid Metals Applications for Fusion (ISLA-8)

The International Symposium on Liquid Metals Applications for Fusion (ISLA) aims to assemble scientists and engineers engaged in research on lithium and liquid metal applications for fusion devices, facilitating discussions on recent advancements and challenges in an open forum to support the development of viable fusion reactors. The 8th International Symposium on Liquid Metals Applications for Fusion (ISLA-8) was organized by the Institute of Plasma Physics, Chinese Academy of Sciences, from 8 to 12 September 2024, in Hefei, China. The symposium was attended by over 70 participants, marking one of the highest attendance figures in the series. A total of 68 presentations were delivered, including 59 presented on-site, while the remainder were conducted online. The participants represented 10 countries, namely China, Japan, the Netherlands, Russia, the USA, the Czech Republic, Italy, Thailand, Spain, and Germany. The symposium covered 10 topics, structured into 13 sessions. Additionally, an opening session provided an overview of the current symposium, while a closing session summarized reports from each session chair.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Steady-State Mechanical Analysis for Target Assembly in the Material Plasma Exposure eXperiment Facility

The Material Plasma Exposure eXperiment (MPEX) steady-state linear plasma facility is currently under design at Oak Ridge National Laboratory to expose target specimens to fusion divertor regimes. The neutron-irradiated target is actively cooled and remote handled in the MPEX facility for conducting plasma-material–interaction (PMI) experiments. Here, the steady-state stresses in the target and target assembly system are investigated using two-dimensional (2-D) and three-dimensional (3-D) models to provide expected stresses/strains under the heat loads to which various system components would be exposed during MPEX operation. The calculated temperatures from the 2-D axisymmetric mechanical model were found to be in excellent agreement with those from the full 3-D thermohydraulic model, providing a strong model validation. Numerical simulation results for the steady-state mechanical model indicate nonuniform distributions for the temperature, stress, and deformation within the critical components. For the initial design, the deformation results indicate possible gap openings between contacting surfaces below the plasma-facing materials. To reduce the possibility of interfacial gap opening, the target assembly was slightly changed and evaluated using the 2-D stress model. Numerical simulation results indicate that the interfacial gap openings can be minimized without drastically changing the entire target assembly. The stress-strain conditions for the target will be further used to assess the appropriate operation during MPEX experiments and gain insight into materials science phenomena during PMI.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Thermohydraulic Design Analysis of the Target Assembly in the Material Plasma Exposure Experiment Facility

The Material Plasma Exposure eXperiment (MPEX) project seeks to design a steady-state linear plasma facility at Oak Ridge National Laboratory that will be used to study plasma-material interactions (PMIs) at fusion prototypic levels, supporting the evaluation and development of materials for the next generation of fusion devices. This study is focused on PMI exposure of small-size neutron-irradiated specimens, which are clamped onto an actively cooled component. A thermohydraulic evaluation of a new MPEX target assembly design to assess the appropriate operation during MPEX operation is presented. To further guide the design and assess the structural integrity of the components under expected loads, preliminary thermomechanical stress analyses were also conducted. To ensure good thermal contact between the components, thermal interface materials, such as silver flexible graphite, were used in the assembly. It was found that the maximum target temperatures of 1572, 1463, and 1315 K were obtained for Grafoil thicknesses of 0.61, 0.38, and 0.25 mm, respectively. The distribution of the axial deformation at high heat fluxes showed that there are no gaps between components, indicating good contact at material interfaces. Moreover, the contact pressure between the target and other components indicated that very good contact was established at these interfaces. Furthermore, the stress-strain conditions for the target will be further used to assess the appropriate operation during MPEX experiments and gain insight into materials science phenomena during PMI experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Investigation of interfacial strength in nacre-mimicking tungsten heavy alloys for nuclear fusion applications

Abstract Tungsten heavy alloys have been proposed as plasma facing material components in nuclear fusion reactors and require experimental investigation in their confirmation. For this purpose, a 90W–7Ni–3Fe alloy has been selected and microstructurally manipulated to present a multiphase brick-and-mortar structure of W-phase ‘bricks’ surrounded by a ductile ‘mortar’. This work draws inspiration from nature to artificially imitate the extraordinary combination of strength and stiffness exhibited by mollusks and produce a nacre-mimicking metal matrix composite capable of withstanding the extremely hostile environment of the reactor interior and maintaining structural integrity. The underlying mechanisms behind this integrity have been probed through high-resolution structural and chemical characterization techniques and have revealed chemically diffuse phase boundaries exhibiting unexpected lattice coherency. These features have been attributed to an increase in the energy required for interfacial decohesion in these systems and the simultaneous expression of high strength and toughness in tungsten heavy alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Thermal diffusivity recovery and defect annealing kinetics of self-ion implanted tungsten probed by insitu transient grating spectroscopy

Tungsten is a promising candidate material for plasma-facing armour components in future fusion reactors. A key concern is irradiation-induced degradation of its normally excellent thermal transport properties. In this comprehensive study, thermal diffusivity degradation in ion-implanted tungsten and its evolution from room temperature (RT) to 1073 K is considered. Five samples were exposed to 20 MeV self-ions at RT to achieve damage levels ranging from 3.2 x 10 -4 to 3.2 displacements per atom (dpa). Transient grating spectroscopy with in-situ heating was then used to study thermal diffusivity evolution as a function of temperature. Using a kinetic theory model, an equivalent point defect density is estimated from the measured thermal diffusivity. The results showed a prominent recovery of thermal diffusivity between 450 K and 650 K, which coincides with the onset of mono-vacancy mobility. After 1073 K annealing samples with initial damage of 3.2 x 10 -3 dpa or less recover close to the pristine value of thermal diffusivity. For doses of 3.2 x 10 -2 dpa or higher, on the other hand, a residual reduction in thermal diffusivity remains even after 1073 K annealing. Transmission electron microscopy reveals that this is associated with extended, irradiation-induced dislocation structures that are retained after annealing. A sensitivity analysis shows that thermal diffusivity provides an efficient tool for the assessing total defect content in tungsten up to 1000 K.

36 MATERIALS SCIENCE↗