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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
NASA NTRS1975

Frontiers in propulsion research: Laser, matter-antimatter, excited helium, energy exchange thermonuclear fusion

Concepts are described that presently appear to have the potential for propulsion applications in the post-1990 era of space technology. The studies are still in progress, and only the current status of investigation is presented. The topics for possible propulsion application are lasers, nuclear fusion, matter-antimatter annihilation, electronically excited helium, energy exchange through the interaction of various fields, laser propagation, and thermonuclear fusion technology.

Papailiou, D. D.↗
NASA NTRS1999

Inertial-Electrostatic Confinement (IEC) Fusion For Space Propulsion

An Inertial-Electrostatic Confinement (IEC) device was assembled at the Marshall Space Flight Center (MSFC) Propulsion Research Center (PRC) to study the possibility of using IEC technology for deep space propulsion and power. Inertial-Electrostatic Confinement is capable of containing a nuclear fusion plasma in a series of virtual potential wells. These wells would substantially increase plasma confinement, possibly leading towards a high-gain, breakthrough fusion device. A one-foot in diameter IEC vessel was borrowed from the Fusion Studies Laboratory at the University of Illinois @ Urbana-Champaign for the summer. This device was used in initial parameterization studies in order to design a larger, actively cooled device for permanent use at the PRC.

Nadler, Jon↗
NASA NTRS1999

Inertial-Electrostatic Confinement (IEC) Fusion for Space Propulsion

An Inertial-Electrostatic Confinement (IEC) device was assembled at the Marshall Space Flight Center (MSFC) Propulsion Research Center (PRC) to study the possibility of using EEC technology for deep space propulsion and power. Inertial-Electrostatic Confinement is capable of containing a nuclear fusion plasma in a series of virtual potential wells. These wells would substantially increase plasma confinement, possibly leading towards a high-gain, breakthrough fusion device. A one-foot in diameter IEC vessel was borrowed from the Fusion Studies Laboratory at the University of Illinois@Urbana-Champaign for the summer. This device was used in initial parameterization studies in order to design a larger, actively cooled device for permanent use at the PRC.

Nadler, Jon↗
NASA NTRS1999

Inertial Electrostatic Confinement Fusion: The Laser Elevator Solar System Survey for Propellants Abstract

Dr. Wayne Pryor worked on three projects this summer. These were: 1) Inertial Electrostatic Confinement; 2) The Laser Elevator; and 3) Solar System Survey for Propellants Abstract. We Assisted Jon Nadler from Richland Community College in assembling and operating a table-top nuclear fusion reactor. We successfully demonstrated neutron production in a deuterium plasma. Pryor also obtained basic spectroscopic information on the atomic and molecular emissions in the plasma. The second project consisted of the completion of a paper on a novel propulsion concept (due to Tom Meyer of Colorado, the first author): a laser sail that bounces light back to the laser source. Recycling the photons from source to sail perhaps 100-1000 times dramatically improves the energy efficiency of this system, which may become very important for high-velocity missions in the future. Lastly, we compiled a very basic inventory of solar system propellant resources, their locations, and their accessibility. This initial inventory concentrates on sunlight availability, water availability, and the difficulty (delta-velocity requirement and radiation environment) in getting there.

Pryor, Wayne↗
DOE OSTI2024

Kinetic Plasma Simulation in the MOOSE Framework: Verification of Electrostatic Particle In Cell Capabilities

In magnetic confinement nuclear fusion reactors, the interaction between the plasma edge and plasma facing components is extremely important. At the plasma edge, a kinetic representation such as particle-in-cell (rather than a fluid representation) is required to accurately capture the plasma behavior. General purpose particle-in-cell plasma simulation capabilities have been developed in the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. This new capability is a part of the development of a new MOOSE-based framework for modeling plasma facing components, the Fusion ENergy Integrated multiphys-X (FENIX) framework. In this work, the verification of foundational particle-in-cell capabilities in FENIX is presented. This new plasma simulation capability has three main components: moving particles in discrete steps on the finite element mesh, mapping charge density from the particle's location to the finite element mesh, and solving for the electrostatic potential based on the charge density mapped from particles to the mesh. In this paper, simple verification problems demonstrating each of these new capabilities are presented, and future work includes electromagnetic capabilities and Monte Carlo collisions with neutral gas particles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2002

Magnetized Target Fusion

Magnetized target fusion (MTF) is under consideration as a means of building a low mass, high specific impulse, and high thrust propulsion system for interplanetary travel. This unique combination is the result of the generation of a high temperature plasma by the nuclear fusion process. This plasma can then be deflected by magnetic fields to provide thrust. Fusion is initiated by a small traction of the energy generated in the magnetic coils due to the plasma's compression of the magnetic field. The power gain from a fusion reaction is such that inefficiencies due to thermal neutrons and coil losses can be overcome. Since the fusion reaction products are directly used for propulsion and the power to initiate the reaction is directly obtained from the thrust generation, no massive power supply for energy conversion is required. The result should be a low engine mass, high specific impulse and high thrust system. The key is to successfully initiate fusion as a proof-of-principle for this application. Currently MSFC is implementing MTF proof-of-principle experiments. This involves many technical details and ancillary investigations. Of these, selected pertinent issues include the properties, orientation and timing of the plasma guns and the convergence and interface development of the "pusher" plasma. Computer simulations of the target plasma's behavior under compression and the convergence and mixing of the gun plasma are under investigation. This work is to focus on the gun characterization and development as it relates to plasma initiation and repeatability.

Steven T. Griffin↗
NASA NTRS1999

Performance Optimization of the Gasdynamic Mirror Propulsion System

Nuclear fusion appears to be a most promising concept for producing extremely high specific impulse rocket engines. Engines such as these would effectively open up the solar system to human exploration and would virtually eliminate launch window restrictions. A preliminary vehicle sizing and mission study was performed based on the conceptual design of a Gasdynamic Mirror (GDM) fusion propulsion system. This study indicated that the potential specific impulse for this engine is approximately 142,000 sec. with about 22,100 N of thrust using a deuterium-tritium fuel cycle. The engine weight inclusive of the power conversion system was optimized around an allowable engine mass of 1500 Mg assuming advanced superconducting magnets and a Field Reversed Configuration (FRC) end plug at the mirrors. The vehicle habitat, lander, and structural weights are based on a NASA Mars mission study which assumes the use of nuclear thermal propulsion' Several manned missions to various planets were analyzed to determine fuel requirements and launch windows. For all fusion propulsion cases studied, the fuel weight remained a minor component of the total system weight regardless of when the missions commenced. In other words, the use of fusion propulsion virtually eliminates all mission window constraints and effectively allows unlimited manned exploration of the entire solar system. It also mitigates the need to have a large space infrastructure which would be required to support the transfer of massive amounts of fuel and supplies to lower a performing spacecraft.

Emrich, William J., Jr.↗
NASA NTRS1990

Speculation on a Solar Chronometer for Climate

Solar activity has been correlated to climatic fluctuations and has been postulated as a major factor in quasi-periodic global climatic change. However, correlations are not explanations of physical mechanisms and do not couple cause with effect. A mechanism for a chronometer for solar output variability is proposed based on relations between properties of thermonuclear fusion, nuclear magnetic moment, and nuclear magnetic resonance. A fundamental oscillation of a nucleus with a net nuclear magnetic moment (NMM) is the precession of its axis of rotation when subjected to a magnetic field. Nuclear magnetic resonance (NMR) is the preferred frequency of precession for a nucleus of a particular isotope when placed in a magnetic field of specific intensity. The NMM for those isotopes involved in the proton-proton (p-p) chain pathway for solar fusion varies from strong positive to strong negative. Individual fusion events, for hydrogen and helium isotopes which release varying amounts of energy, may be controlled by NMR frequencies. The pulses of energy from fusion events occurring at NMR frequencies in the solar interior may be transformed into pressure or gravity waves that emerge as gravity or acoustic waves at the surface. Dictated by spherical harmonics, certain wavelengths may be reinforced and reenter the solar interior to modulate the fusion process. Qualitative analysis of solar and climatic data support the interaction of the three basic components of the chronometer, magnetic activity, oscillation frequency, and solar energy output.

Perry, Charles A.↗
NASA NTRS2007

Muon Catalyzed Fusion

Muon catalyzed fusion is a process in which a negatively charged muon combines with two nuclei of isotopes of hydrogen, e.g, a proton and a deuteron or a deuteron and a triton, to form a muonic molecular ion in which the binding is so tight that nuclear fusion occurs. The muon is normally released after fusion has taken place and so can catalyze further fusions. As the muon has a mean lifetime of 2.2 microseconds, this is the maximum period over which a muon can participate in this process. This article gives an outline of the history of muon catalyzed fusion from 1947, when it was first realised that such a process might occur, to the present day. It includes a description of the contribution that Drachrnan has made to the theory of muon catalyzed fusion and the influence this has had on the author's research.

Armour, Edward A.G.↗
NASA NTRS2015

Lightweight Damage Tolerant, High-Temperature Radiators for Nuclear Power and Propulsion

NASA is increasingly emphasizing exploration to bodies beyond near-Earth orbit. New propulsion systems and new spacecraft are being built for these missions. As the target bodies get further out from Earth, high energy density systems, e.g., nuclear fusion, for propulsion and power will be advantageous. The mass and size of these systems, including supporting systems such as the heat exchange system, including thermal radiators, will need to be as small as possible. Conventional heat exchange systems are a significant portion of the total thermal management mass and size. Nuclear electric propulsion (NEP) is a promising option for high-speed, in-space travel due to the high energy density of nuclear fission power sources and efficient electric thrusters. Heat from the reactor is converted to power for use in propulsion or for system power. The heat not used in the power conversion is then radiated to space as shown in figure 1. Advanced power conversion technologies will require high operating temperatures and would benefit from lightweight radiator materials. Radiator performance dictates power output for nuclear electric propulsion systems. Pitch-based carbon fiber materials have the potential to offer significant improvements in operating temperature, thermal conductivity, and mass. These properties combine to allow significant decreases in the total mass of the radiators and significant increases in the operating temperature of the fins. A Center-funded project at NASA Marshall Space Flight Center has shown that high thermal conductivity, woven carbon fiber fins with no matrix material, can be used to dissipate waste heat from NEP systems and because of high specific power (kW/kg), will require less mass and possibly less total area than standard metal and composite radiator fins for radiating the same amount of heat. This project uses an innovative approach to reduce the mass and size required for the thermal radiators to the point that in-space NEP and power is enabled. High thermal conductivity carbon fibers are lightweight, damage tolerant, and can be heated to high temperature. Areal densities in the NASA set target range of 2 to 4 kg/m2 (for enabling NEP) are achieved and with specific powers (kW/kg) a factor of about 7 greater than conventional metal fins and about 1.5 greater than carbon composite fins. Figure 2 shows one fin under test. All tests were done under vacuum conditions.

Craven, Paul D.↗
DOE OSTI2025

First Demonstration of Improved Fusion Yield with Increased Compression through Reduced Adiabat in Inertial Confinement Fusion Experiments at the National Ignition Facility

Recent advancements in indirect-drive inertial confinement fusion (ICF) experiments at the National Ignition Facility (NIF) have achieved a significant milestone by demonstrating target gains greater than one, yet future applications necessitate much higher target gains. One approach to achieving improved implosion performance is to pursue increased fuel compression via a lowered implosion adiabat. Experiments have been performed testing a reduced adiabat by introducing small changes to the drive laser pulse shape and the resulting shock timing for an existing implosion design at 1.9 MJ laser drive with near-ignition performance (experiment N210808). Experiments using the updated design demonstrate, for the very first time, increased compression and fusion yield in ICF implosions on the NIF by using a lower fuel adiabat, and increased compression with a reduced adiabat in high-density carbon ablators. Compared to the previously best-performing experiment with a laser energy of 1.9 MJ, these experiments exhibit increases of up to 80% and 14% in nuclear fusion yield and fuel compression, respectively, and with repeatable performance. Further, it is the only implosion design to have achieved a target gain exceeding one with a laser energy of less than 2 MJ. These findings highlight the efficacy of reduced adiabat designs in achieving higher compression and fusion yields, offering a promising pathway for future ICF applications. In conclusion, this Letter not only addresses a long-standing question in ICF but also paves the way for achieving higher target gains with optimized implosion strategies.

Hohenberger, M. [Lawrence Livermore National Labor↗
NASA NTRS1992

Fusion energy for space: Feasibility demonstration. A proposal to NASA

This proposed program is to initiate a space flight research and development program to develop fusion energy for the space applications of direct space propulsion and direct space power, that is, a Space Fusion Energy (SFE) program. 'Direct propulsion' refers to the use of plasma energy directly for thrust without requiring other energy conversion systems. Further, to provide space missions with large electrical power, 'direct space power' is proposed whereby the direct conversion of charged particles into electricity is used, thereby avoiding thermal conversion system losses. The energy release from nuclear fusion reactions makes these highly efficient, high power space systems possible. The program as presented conducts in an orderly, hierarchical manner the necessary planning, analyses, and testing to demonstrate the practical use of fusion energy for space. There is nothing discussed that is known to be theoretically impossible. Validation of the engineering principles is sought in this program which uses a cost-benefit approach. Upon successful program completion, space will become more accessible and space missions more safely conducted. The country will have taken a giant step toward the commercialization of space. The mission enabling capability provided by fusion energy is well beyond mission planners' current dreams.

Schulze, Norman R.↗
DOE OSTI2025

Laser Powder Bed Fusion Additive Manufacture Nb1Zr Development

Next generation fission and fusion nuclear reactors require materials that can withstand operating temperatures greater than 500 °C, neutron irradiation doses of up to 200 displacements per atom (dpa), and potentially corrosive coolants such as the alkali liquid metals sodium, lithium, and NaK (Na33K eutectic alloy). Refractory alloys, such as Nb1Zr (Nb-1wt%Zr) and Molybdenum alloy TZM (Mo-0.5wt%Ti-0.08wt%Zr) have been traditionally considered viable candidates for advanced fission and fusion reactor concepts. However, it is relatively difficult to generate complex geometries of interest from these alloys using traditional manufacturing methods. In addition, there needs to be a concentrated effort to address refractory metal challenges at elevated temperature operation. In order to generate complex geometries of interest, modern manufacturing techniques are considered to increase the technological readiness level (TRL), cost-effectiveness, and schedule savings. This work focused on the continued development of laser powder bed fusion (L-PBF) additive manufacturing (AM) to improve both design flexibility, evaluate microstructure and properties, and ultimately accelerate the TRL and qualification of these processes and alloys for components to potentially be put into service. Niobium alloy Nb1Zr was identified through a down-selection process outlined in previous reports as a candidate to develop in L-PBF AM. Historically, Nb1Zr had been explored for high temperature fast spectrum fission reactors for both terrestrial and space applications. Molybdenum alloy TZM has also been considered for these reactor concepts due to exceptional high-temperature strength, creep resistance, and stability under irradiation. L-PBF AM of TZM has previously been investigated at LANL under the Microreactor program, NASA, ORNL, and in academia. However, due to the crack prone nature of TZM, L-PBF AM of TZM resulted in significant microcracking and additional development is required to pursue viable maturation. Other AM methods have been found to be more successful in printing TZM, and those alternatives approaches are discussed in this effort. The efforts detailed in this report focused on continued development of Nb1Zr through L-PBF and development of TZM via L-PBF and electron powder bed fusion (E-PBF). The objective of this work was to further the development of these AM techniques for the chosen refractory alloys, elucidating and addressing associated challenges through characterization of several demonstration builds. At LANL, Nb1Zr builds were completed using an EOS M290 and M400 machines, and a refractory alloy-dedicated L-PBF system, the Xact Metal XM200G, was installed. The XM200G primary purpose was to do the Nb1Zr parameter development process; however, due to difficulties associated with the machine installation and qualification process, it was decided to pivot development to the larger M400 and M290 machines. Although the supply of Nb1Zr powder was limited, it was sufficient to generate sub-scale metallographic specimens for the purpose of parameter development. This was first accomplished on the EOS M400 then the M290 due to machine schedule availability. Further development of TZM has been initiated at the University of Texas El Paso (UTEP) under contract with LANL to use both a heated build envelope L-PBF machine and E-PBF machine that have been found in the literature to mitigate microcracking. UTEP was provided with TZM powder and build plates to support parallel TZM parameter development across both machines. As part of the contract, UTEP will also be conducting microstructural characterization once optimized process parameters have been identified. The optimized process parameters for each machine will be used to generate a series of metallographic, mechanical, and surface finish specimens for subsequent characterization and testing. In the next section, we provide a detailed discussion of the methodology used for investigating the feasibility of leveraging these alloys for use in advanced reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

The Maintainable Fusion Pilot Plant

The US fusion community has coalesced around the goal of building an FPP as described by the National Academies of Science, Engineering, and Medicine (NASEM). In addition to demonstrating the viability of the technologies necessary to operate such a plant, including demonstration of net energy and electricity production, NASEM found that a “fusion pilot plant will need to demonstrate the ability to efficiently perform remote maintenance and replacement in support of the design of a power plant, taking into account details of the consequences of the fusion environment, such as material activation and tritium retention in components.” Current designs of fusion demonstration reactors do usually foresee a regular exchange of their first wall modules, including the tritium breeding blankets. In the European Power Plant Conceptual Studies, it is assumed that a fusion reactor will need to change its divertor every 2 years and its first wall blanket module every 5 to 6 years to reach acceptable availability. Underlying this capability are remote-handling technologies to keep the outage for the exchange of these components short. There are many uncertainties in the remote-handling schemes, and most schemes are at a preconceptual level at best. In addition, the exchange of these components would either produce an enormous rad-waste stream or would require an enormous refurbishment activity with huge cost-prohibitive hot-cells. Past Fusion Nuclear Science Facility (FNSF) preconceptual studies have led to hot cell dimensions of an unbelievable size, likely costing tens of billions of dollars. Already at The Way (previously International Thermonuclear Experimental Reactor, ITER), hot-cells have become cost-prohibitive, demanding redesigns of the ITER first wall to reduce the toxic rad-waste/inventory. In this in-situ PFC repair project, a concept for a long-life, maintainable first wall module concept is developed and tested. This first wall concept relies on innovative remote handling to repair the first wall modules in-situ, avoiding costly refurbishments outside of the tokamak vessel. This approach was highlighted in the Fusion Energy Sciences Advisory Committee (FESAC) report on Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy. In general, the damage of the first wall armor is due to particle and radiation exposures. Load conditions vary from one fusion reactor design to another. In tokamaks, first wall Plasma Facing Components (PFCs) are exposed to far-Scrape-Off-Layer plasma fluxes, electromagnetic radiation, energetic CX neutrals, and potentially runaway electron beams. Protecting the first wall to the worst-case load conditions would require the design of a very thick first wall armor. Transient heat and particle fluxes due to disruptions or edge localized modes can lead to excessive heat loads resulting potentially in melting PFCs down to the cooling channel. Catastrophic events like these need to be avoided by appropriate disruption mitigation systems. However, failure of these systems will still put a first wall at an unacceptable risk. Hence, a first wall design needs to accommodate the occasional transient heat loads by introducing sacrificial limiters, which will absorb these transients.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Manufacturing Li 2 TiO 3 -based tritium breeder materials by volume-controlled spark plasma sintering with an optimized microstructure

Multifunctional ceramic breeder materials are highly desirable for the deuterium-tritium fusion to achieve high efficiency in breeding tritium through neutron irradiation of lithium-containing blankets. Li 2 TiO 3 displays unique attributes as a potential ceramic breeder material. An optimized microstructure with three-dimensional interconnected pore structure is required for rapid transport of the tritium for effective fuel cycle, which however enviably results in the degradation of the thermal-mechanical properties of the breeding materials. In this work, nanocrystalline porous Li 2 TiO 3 ceramic pellets with controlled porosities of 14% and 20% are manufactured by volume-controlled spark plasma sintering. An optimized 3D interconnected pore structure is achieved consisting of both micro-sized pores and nano-sized pores embedded in nanocrystalline matrix, which could be beneficial to facilitate easy removal of bred T and He. Further, the 3D interconnected porous structure is well maintained upon isothermal annealing of the SPS-fabricated pellets at relevant operation temperature of the solid breeding materials. Single-phasic porous pellets also display enhanced thermal-mechanical properties, superior to current state-of-the-art materials which establish their potential as a promising tritium breeder material for nuclear fusion applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Suppression of irradiation hardening in tungsten-coated ferritic steel for fusion reactor blanket applications

W-coated reduced activation ferritic steels have been developed for use as plasma facing components in fusion reactor blankets, offering excellent sputtering resistance and structural strength. Previous high-temperature coating methods, such as diffusion bonding and brazing, caused interfacial deterioration due to thermal stress from mismatched thermal expansion between W and reduced activation ferritic steel. To address this, underwater explosive welding was introduced as a high-velocity cold process that joins dissimilar materials while maintaining a strong, thin interface without the thermal issues associated with traditional methods. In this study, the effects of neutron irradiation on the hardness and microstructure in W-coated F82H reduced activation ferritic steel (W/F82H) joined by underwater explosive welding are investigated. Following neutron irradiation at 290 °C, irradiation hardening is suppressed in W, F82H, and their interface within the W/F82H material. Furthermore, microstructural observations indicate that the recovery of work hardening and relaxation of elastic strain introduced during coating significantly contribute to the suppression of irradiation hardening in W/F82H. In conclusion, W/F82H exhibits significantly suppressed irradiation hardening compared with those in stand-alone materials. This suppression is explained by residual stress from thermal expansion mismatch and the unique microstructure at the interface. These results provide valuable insights for the development of more durable materials in nuclear fusion applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Liquid metal walls

Here, the plasma performance of fusion devices depends strongly on the chosen wall materials. Solid plasma-facing components (PFCs) are predominantly used in present devices, and are the most investigated candidates for fusion designs. High-Z materials such as tungsten (W) are the leading solid PFC material candidates. To date, a material choice that scales to steady-state reactor conditions has not been identified. Moreover, if plasma transient events such as edge-localized modes and disruptions cannot be altogether avoided or sufficiently mitigated, the projected peak heat and particle loads far exceed the power exhaust capabilities of solids. Liquid metal (LM) PFCs represent an intrinsic self-healing boundary that are both resilient to surface damage from transients, and that could handle high steady-state heat and particle fluxes. Flowing LM PFCs can be designed to remove “slag,” the buildup of material erosion due to plasma-material interactions, including charge exchange sputtering in the main chamber. Further, LM offer the prospect to manage hydrogenic species otherwise retained in the PFCs, which is important from a safety and inventory standpoint. The two most promising LMs are lithium (Li) and tin (Sn), although Sn–Li eutectics may be considered. While Sn offers a higher temperature window with low vapor pressure and low hydrogen retention, Li offers the prospect of enhanced energy confinement and higher acceptable core contamination limits, and this section focuses on Li PFCs. An LM PFC development research program developed LM PFC concepts for a nuclear fusion device via engineering design calculations, single-effect experiments, and staged prototypical experiments. A self-consistent design window was identified with liquid Li flow speeds ~5–10 m/s; plasma contamination was negligible for predicted Li evolution rates. While these preconceptual designs hold promise, there is substantial R&D needed to advance the technical readiness levels of LM PFCs for application to fusion power plants.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Pressureless sintering of lithium hydride

Lithium Hydride is a material of growing importance for addressing technological challenges related to nuclear fusion, long-term human space travel, and thermal energy storage. Pressureless sintering provides a straightforward, scalable approach to produce dense LiH parts of all sizes and shapes. Pressed LiH green compacts were sintered at heating rates from 2.5 to 20 °C/min to 650 °C, yielding densities up to 96 ± 1.4 %, with densification initiating at 500 °C. Here, a validated master sintering curve was constructed with a sintering apparent activation energy of 135 kJ/mol. X-ray diffraction and simultaneous thermal analysis revealed Li 2 O formation from 300 – 550 °C and decomposition of LiH into Li metal at 550 °C, each reflected as deviations in the master sintering curve. Computed tomography after thermal treatment to 550 °C showed the formation of corrosion products, and after thermal treatment to 650 °C LiH reduction to Li most significantly at exposed surfaces.

36 MATERIALS SCIENCE↗