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DOE OSTI2023

Thermal quench induced by a composite pellet-produced plasmoid

Injecting shattered pellets is the critical concept of the envisaged ITER disruption mitigation system (DMS). Rapid deposition of large amounts of material should presumably result in controlled cooling of the entire plasma. A considerable transfer of thermal energy from the electrons of the background plasma to the ions accompanies a localized material injection due to the ambipolar expansion along the magnetic field line of the cold and dense plasmoid produced by the ablated pellet. Radiation initially plays the dominant role in the energy balance of a composite plasmoid containing high-Z impurities. A competition between the ambipolar expansion and the radiative losses defines the Thermal Quench scenario, including the amount of pre-quench thermal energy radiated on a short collisional timescale—possibly detrimental for the plasma-facing components. The present work quantifies plasmoid energy balance for disruption mitigation parameters. For pure hydrogen injection, up to 90% of the pre-pellet electron thermal energy may go to the newly injected ions. We also demonstrate that a moderate high-Z impurity content within the plasmoid can reduce highly localized radiation at the beginning of the expansion. The thermal energy will then dissipate on the much longer ion collisional timescale, which would be attractive for ITER DMS.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Impact of Soret effect on hydrogen and helium retention in PFC tungsten under ELM-like conditions

Abstract In our previous work, we have demonstrated using nonequilibrium molecular-dynamics simulations that the fluxes of helium and self-interstitial atoms in the presence of a thermal gradient in tungsten are directed opposite to the heat flux, indicating that species transport is governed by a Soret effect, namely, thermal-gradient-driven diffusion, characterized by a negative heat of transport that drives species transport uphill, i.e. from the cooler to the hot regions of the tungsten sample. In this work, the findings of our thermal and species transport analysis have been implemented in our cluster-dynamics code, Xolotl, which has been used to compute temperature and species profiles over spatiotemporal scales representative of plasma-facing component (PFC) tungsten under typical reactor operating conditions, including extreme heat loads at the plasma-facing surface characteristic of plasma instabilities that induce edge localized modes (ELMs). We demonstrate that the steady-state species profiles, when properly accounting for the Soret effect, vary significantly from those where temperature-gradient-driven transport is not accounted for and discuss the implications of such a Soret effect on the response to plasma exposure of plasma-facing tungsten. Although our cluster-dynamics simulations do not yet include self-clustering of helium or hydrogen blister formation, our simulation results show that the Soret effect substantially reduces helium and hydrogenic species retention inside PFC tungsten.

36 MATERIALS SCIENCE↗

DOE OSTI2026

Multiphysics Analysis of Li Cooled Divertor Substrate During Loss of Coolant Accident (LOCA)

In the ongoing study of potential designs for liquid metal (LM) plasma-facing components (PFCs), so-called “slow” and “fast” Li flow divertor concepts are under investigation. In the previous studies on design and analysis of the slow Li flow divertor and comparison with the fast Li flow divertor, the magnetohydrodynamics (MHD)/heat transfer effects of the Li flowing inside the substrate as a second coolant were comprehensively investigated under the normal steady-state operation conditions. Here, in the present study, the multiphysics analysis is extended to the unsteady abnormal divertor scenario where the Li layer on top of the substrate does not provide full coverage or even totally disappears for a certain period of time, so that the substrate becomes directly exposed to the incident high plasma heat flux. Such an unwanted event may happen regardless of the concept of the divertor and is worth detailed investigations, typically referred to as a loss of coolant accident (LOCA). To address this situation, a simplified scoping analysis is conducted first in 2-D, and then an integrated 3-D modeling is performed using a time-dependent multiphysics model in COMSOL Multiphysics that integrates LM MHD, heat transfer, and solid mechanics. The main goal is to evaluate conditions under which the major material limits, such as the maximum allowable temperature, stress, and displacement of the substrate, can still be met. It was shown that the maximum time over which the substrate of RAFM steel can retain structural integrity during the LOCA is around 0.2 ∼ 0.3 s. Any divertor concept that utilizes RAFM steel as a substrate material and liquid Li as a second coolant should take such a permitted time into consideration.

divertor↗

DOE OSTI2021

Effect of boronization on plasma-facing graphite surfaces and its correlation with the plasma behavior in NSTX-U

Boronization is a Plasma Facing Component (PFC) conditioning technique widely used in tokamak machines. The National Spherical Torus Experiment-Upgrade (NSTX-U) applied this conditioning, using a plasma glow with a deuterated Trimethyl-boron (d-TMB) and He mixture. The use of boronization during the campaign improved the plasma performance, allowing longer plasma discharges and H-mode access. The chemical state of an ATJ graphite sample, used as a proxy for the NSTX-U PFCs, was monitored in-situ using the Materials Analysis Particle Probe (MAPP) diagnostic and X-ray Photoelectron Spectroscopy (XPS). The XPS data showed a progressive rise (from + fluence increased. Filterscopes were used to measure the light emitted by oxygen impurities in the plasma near the surface of the PFC. An increase in the registered magnitude of the OII line, normalized to the Dγ intensity, was observed as the concentration of O on the ATJ surface increased. The plasma performance was found to be strongly correlated to oxygen impurity concentrations at the plasma edge and on the PFC surface, as measured by the discharge length and access to the H-mode regime. In this work, we present a quantitative analysis of the evolution of the chemistry of the ATJ surface, and the oxygen presence in the plasma-material interface, and report relevant plasma parameters observed during the same period of time.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

NASA NTRS1973

Pulsed plasma microthruster for synchronous meteorological satellite /SMS/

Two completely self contained variable thrust flight prototype plasma propulsion systems using solid teflon as propellant were designed, built and laboratory tested. The impulse bit amplitude of each thruster is 25 micro lb-sec. Impulse bits can be provided at rates varying from 50 ppm to 110 ppm. The total impulse capability is 400 lb-sec. Only 23 watts of bus power are required at maximum pulse rate. The system was subjected and passed the required vibration schedule and thermal vacuum test after a minor change in capacitor end face material. A 13 million discharge life test of the system was subsequently performed.

Guman, W. J.↗

DOE OSTI2025

Roadmap for warm dense matter physics

This roadmap presents the state-of-the-art, current challenges and near future developments anticipated in the thriving field of warm dense matter physics. Originating from strongly coupled plasma physics, high pressure physics and high energy density science, the warm dense matter physics community has recently taken a giant leap forward. This is due to spectacular developments in laser technology, diagnostic capabilities, and computer simulation techniques. Only in the last decade has it become possible to perform accurate enough simulations \& experiments to truly verify theoretical results as well as to reliably design experiments based on predictions. Consequently, this roadmap discusses recent developments and contemporary challenges that are faced by theoretical methods, and experimental techniques needed to create and diagnose warm dense matter. A large part of this roadmap is dedicated to specific warm dense matter systems and applications in astrophysics, inertial confinement fusion and novel material synthesis.

FOS: Physical sciences↗

DOE OSTI2024

Renewable low-Z wall for fusion reactors with built-in tritium recovery (Final Technical Report)

This project pursued development of a novel renewable plasma-facing wall technology for fusion reactors. The technology is based on a slurry which can be easily delivered by delivery tubes to the reactor wall. The slurry dries at the hot reactor wall into pebble rods which are extruded out into the hot plasma, where the pebbles break off and fall along the reactor wall and can be recovered by gravity and re-used. The falling pebbles carry away heat and tritium and also protect the wall against large scale erosion or redeposition of material. The research focused on carbon-based pebble rods and demonstrated that pebble rods could be produced from slurry with tolerable levels of outgassing on a reactor-relevant timescale (< 5 minutes). Steady-state handling of reactor relevant (up to 50 MW/m 2 ) normal-incidence heat loads was demonstrated. Pebble release velocities were found to be sufficiently small (< 1 m/s) to allow recovery below the vacuum chamber. Tunability of the pebble rod breaking rate was demonstrated by changing the fill fraction of the interpebble matrix which binds the pebbles together. This work could benefit the public by helping move forward the design of commercially viable fusion energy reactors. Designing a first wall for magnetic fusion reactors which can handle the huge heat loads present and also avoid buildup of tritium-containing deposits is extremely challenging and requires novel approaches like the one being investigated here.

36 MATERIALS SCIENCE↗

NASA NTRS1994

Preliminary far-field plume sputtering characterization of the Stationary Plasma Thruster (SPT-100)

For electric propulsion devices to be considered for use on communications satellites, integration impacts must be examined in detail. Two phenomena of concern associated with highly energetic plumes are contamination via sputtered material from the thruster and sputter erosion of downstream surfaces. In order to characterize the net effect of both phenomena, an array of witness plates were mounted in several types of holders and were exposed to the SPT-100 thruster plume for 50 hours. Surface analysis of the witness plates revealed that in the most energetic regions of the plume, there was a net removal of material from the samples facing the thruster. In the peripheral regions, net deposits were observed and characterized by the changes in optical properties of these samples. Changes in surface properties of samples located in collimators were within experimental uncertainty.

Pencil, Eric J.↗

DOE OSTI2025

Comparison of plasma start-up with high Z and low Z first wall in WEST

The choice of first wall material is of paramount importance for the plasma start-up conditions in ITER and future fusion power plants. In this context, the present work focuses on the correlations between first wall impurity sources and total radiated power during plasma start-up in the tungsten (W) Environment in Steady-state Tokamak (WEST). The objective is to highlight experimental indications for a preferable combination of start-up plasma scenario and first wall materials. Until 2019, WEST featured a full high Z first wall with all limiters exposing only W surfaces to the plasma. To study the impact of a low Z first wall in WEST, boron nitride tiles were installed in the central part of the inner and outer limiters in 2020. Although visible spectroscopy and bolometry measurements show respectively a strong weakening of the WI line intensity on the limiters and a reduction of radiated power after the changeover, a degradation occurs with the accumulation of plasma exposure. In addition, the different plasma facing elements of the main chamber do not influence equally the radiated power during start-up. In both high Z and low Z environments, a clear non-linear dependence is found between the start-up radiated power and the outer limiter W impurity source. Since W seems to be the main cause for core radiation, correlation between outer limiter W sources and other impurity sources are investigated. Finally, analysis of the legacy of B powder drops on a number of start-up plasmas suggests that it is less effective at reducing radiated power when the first wall is covered with W.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Conceptual Design and Analysis of In-Vessel Components for the Materials Plasma Exposure eXperiment (MPEX)

The materials plasma exposure experiment (MPEX) is a linear plasma divertor simulator currently undergoing conceptual design. The facility will expose material samples to steady-state plasma fluxes to examine plasma–material interactions (PMIs) that are expected in the next generation of fusion devices. The plasmas will be generated by a helicon source, with electron and ion heating sources of up to 800 kW possible. The peak heat fluxes of the target are expected to be up to 10 MW/m 2 . The facility will be capable of handling low-activation neutron-irradiated samples in order to examine the multivariate effects of neutron damage and plasma fluence. Neutron-irradiated samples are planned to be roughly of 10-mm diameter; however, plasma-facing components up to 60×600 mm can be accommodated. The steady-state nature of the device will require the magnetic confinement of the plasma to be achieved with superconducting magnets, with a maximum on-axis field of 2.5 T. In addition, since MPEX will be a steady-state device, in-vessel components need to be water cooled. The primary in-vessel components will be the target, the dump plate, the limiter, the skimmers, and the microwave absorber. The conceptual design of these components is presented here, including analyses that confirm that the designs are adequate to meet the requirements of MPEX operation.

Lumsdaine, Arnold↗

NASA NTRS1994

New Trends in Coatings Developments for Turbine Blades: Materials Processing and Repair

Turbine engines for aeronautic applications now have to face strenuous requirements concerning not only operating performances but also reliability and repairability. This evolution of the specifications induces important consequences regarding the choice and the design of turbine blade protective coatings. This paper presents several key features concerning the deposition techniques and the real-life behavior of such coatings: (1) complex aluminides well suited to the protection against high temperature oxidation and hot corrosion of directionally solidified nickel-base superalloy turbine blades; (2) ceramic coatings used as thermal barriers - they have to exhibit both high resistance to thermomechanical fatigue and adequate smoothness; (3) processing techniques for thermal barriers, including plasma spraying, electron beam physical vapor deposition and plasma enhanced chemical vapor deposition; and (4) blades and coatings repair techniques.

S Alpérine↗

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↗

NASA NTRS1981

Regenerator seal

A method for manufacturing a hot side regenerator cross arm seal assembly having a thermally stablilized wear coating with a substantially flat wear surface thereon to seal between low pressure and high pressure passages to and from the hot inboard side of a rotary regenerator matrix includes the steps of forming a flat cross arm substrate member of high nickel alloy steel; fixedly securing the side edges of the substrate member to a holding fixture with a concave surface thereacross to maintain the substrate member to a slightly bent configuration on the fixture surface between the opposite ends of the substrate member to produce prestress therein; applying coating layers on the substrate member including a wear coating of plasma sprayed nickel oxide/calcium flouride material to define a wear surface of slightly concave form across the restrained substrate member between the free ends thereon; and thereafter subjecting the substrate member and the coating thereon to a heat treatment of 1600.degree. F. for sixteen hours to produce heat stabilizing growth in the coating layers on the substrate member and to produce a thermally induced growth stress in the wear surface that substantially equalizes the prestress in the substrate whereby when the cross arm is removed from the fixture surface following the heat treatment step a wear face is formed on the cross arm assembly that will be substantially flat between the ends.

Davis, Leonard C.↗

DOE OSTI2026

A Comparative Study of Thermal Oxidization Resistance of a High-Entropy Metal Boride and a High-Entropy Metal Carbide

We present a systematic study of thermal oxidation resistance of transition metal borides and carbides up to 1300 °C in a dry air environment. A High-Entropy Metal Boride (HEMB), of composition (Hf 0.2 , Mo 0.2 , Nb 0.2 , Ta 0.2 , Zr 0.2 )B 2 , and a similar High-Entropy Metal Carbide (HEMC) (Hf, Mo, Nb, Ta, Zr)C 5 were synthesized from precursor mixtures, under 30 MPa of pressure at a temperature of 1800 °C using a Spark Plasma Sintering Device. The synthesized phases were confirmed via X-ray Diffraction analysis, which showed a pure hexagonal AlB 2 -type structure for HEMB and a face-centered cubic (FCC) structure for HEMC, with lattice parameters, a = 3.10 Å and c = 3.37 Å for HEMB and a = 4.524 Å for HEMC. Oxidation resistance was evaluated using a simultaneous thermogravimetric analysis and differential scanning calorimetry (TGA/DSC) stage in which HEMB and HEMC were heated up to 1300 °C at a rate of 2 °C/min in a dry air environment. Scanning electron microscopy (SEM) was used to analyze the resulting oxidized material. Our study demonstrates that HEMB shows better thermal oxidation resistance as compared to a similar metal composition HEMC at high temperatures.

36 MATERIALS SCIENCE↗

NASA NTRS2011

Optimization of Indium Bump Morphology for Improved Flip Chip Devices

Flip-chip hybridization, also known as bump bonding, is a packaging technique for microelectronic devices that directly connects an active element or detector to a substrate readout face-to-face, eliminating the need for wire bonding. In order to make conductive links between the two parts, a solder material is used between the bond pads on each side. Solder bumps, composed of indium metal, are typically deposited by thermal evaporation onto the active regions of the device and substrate. While indium bump technology has been a part of the electronic interconnect process field for many years and has been extensively employed in the infrared imager industry, obtaining a reliable, high-yield process for high-density patterns of bumps can be quite difficult. Under the right conditions, a moderate hydrogen plasma exposure can raise the temperature of the indium bump to the point where it can flow. This flow can result in a desirable shape where indium will efficiently wet the metal contact pad to provide good electrical contact to the underlying readout or imager circuit. However, it is extremely important to carefully control this process as the intensity of the hydrogen plasma treatment dramatically affects the indium bump morphology. To ensure the fine-tuning of this reflow process, it is necessary to have realtime feedback on the status of the bumps. With an appropriately placed viewport in a plasma chamber, one can image a small field (a square of approximately 5 millimeters on each side) of the bumps (10-20 microns in size) during the hydrogen plasma reflow process. By monitoring the shape of the bumps in real time using a video camera mounted to a telescoping 12 magnifying zoom lens and associated optical elements, an engineer can precisely determine when the reflow of the bumps has occurred, and can shut off the plasma before evaporation or de-wetting takes place.

Jones, Todd J.↗

NASA NTRS1990

Study of diamond film growth and properties

The objective was to study diamond film growth and its properties in order to enhance the laser damage threshold of substrate materials. Calculations were performed to evaluate laser induced thermal stress parameter, R(sub T) of diamond. It is found that diamond has several orders of magnitude higher in value for R(sub T) compared to other materials. Thus, the laser induced damage threshold (LIDT) of diamond is much higher. Diamond films were grown using a microwave plasma enhanced chemical vapor deposition (MPECVD) system at various conditions of gas composition, pressure, temperature, and substrate materials. A 0.5 percent CH4 in H2 at 20 torr were ideal conditions for growing of high quality diamond films on substrates maintained at 900 C. The diamond films were polycrystalline which were characterized by scanning electron microscopy (SEM) and Raman scattering spectroscopy. The top surface of the growing film is always rough due to the facets of polycrystalline film while the back surface of the film replicates the substrate surface. An analytical model based on two dimensional periodic heat flow was developed to calculate the effective in-plane (face parallel) diffusivity of a two layer system. The effective diffusivity of diamond/silicon samples was measured using a laser pulse technique. The thermal conductivity of the films was measured to be 13.5 W/cm K, which is better than that of a type Ia natural diamond. Laser induced damage experiments were performed on bare Si substrates, diamond film coated Si, and diamond film windows. Significant improvements in the LIDT were obtained for diamond film coated Si compared to the bare Si.

Albin, Sacharial↗

DOE OSTI2021

PPPL Laboratory Directed Research and Development (Project Final Reports, FY2018 - FY2020)

The U.S. Department of Energy’s (DOE) Princeton Plasma Physics Laboratory (PPPL) is a collaborative national center for fusion energy science, basic sciences, and advanced technology. The Laboratory has three major missions: (1) to develop the scientific knowledge and advanced engineering to enable fusion to power the U.S. and the world; (2) to advance the science of nanoscale fabrication for technologies of tomorrow; and (3) to further the development of the scientific understanding of the plasma universe from laboratory to astrophysical scales. PPPL’s Laboratory Directed Research and Development (LDRD) program supports and encourages creativity and innovation and contributes to its long-term viability. New scientific and technical research areas emerge and are nurtured through the program. Furthermore, new capabilities are developed to enable the Laboratory to meet its and DOE’s missions. The program is used to systematically diversify the Laboratory’s programs and mission. In the last few years, the program has started projects in nanomaterial synthesis, microelectronics, advanced x-ray spectroscopy, high-energy-density physics, superconducting magnet technology, machine learning and artificial intelligence, 3D magnetic fields to optimize fusion plasmas, integration of permanent magnets with simple high-field magnets to reduce the cost of producing complex 3D magnetic fields, advanced computational methods for predictive understanding and control of fusion plasma, development of quantum computing algorithms for plasma physics, liquid metal plasma-facing components for fusion reactors, virtual engineering, and plasma-based space propulsion. The program is also the vehicle to recruit and train talented scientists and engineers with the new skills needed to perform the Laboratory’s mission. Many of the new hires through the program go on to become world-class scientists and engineers in their fields. This report provides descriptions and accomplishments of those LDRD projects that were completed during fiscal years 2018 through 2020.

36 MATERIALS SCIENCE↗

NASA NTRS2005

Characterization of an 8-cm Diameter Ion Source System

Results of tests characterizing an 8-cm diameter ion source are presented. The tests were conducted in three separate vacuum test facilities at the University of Alabama-Huntsville, Colorado State University, and L3 Communications' ETI division. Standard ion optics tests describing electron backstreaming and total-voltage-limited impingement current behavior as a function of beam current were used as guidelines for selecting operating conditions where more detailed ion beam measurements were performed. The ion beam was profiled using an in-vacuum actuating probe system to determine the total ion current density and the ion charge state distribution variation across the face of the ion source. Both current density and ExB probes were utilized. The ion current density data were used to obtain integrated beam current, beam flatness parameters, and general beam profile shapes. The ExB probe data were used to determine the ratio of doubly to singly charged ion current. The ion beam profile tests were performed at over six different operating points that spanned the expected operating range of the DAWN thrusters being developed at L3. The characterization tests described herein reveal that the 8-cm ion source is suitable for use in (a) validating plasma diagnostic equipment, (b) xenon ion sputtering and etching studies of spacecraft materials, (c) plasma physics research, and (d) the study of ion thruster optics at varying conditions.

Li, Zhongmin↗