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At least 973 records · Page 54

Shape Memory Alloy Material Maturation (SMA)

The Shape Memory Alloy (SMA) project will evaluate, process, and refine 3 SMA materials that have the potential to provide the properties needed to enable various technologies for human exploration missions on the Moon and Mars. The work will focus on the development and identification of a new SMA alloy targeted at enabling compliant rover tires capable of the extreme demands (high loads, faster drive speeds, extreme cold temperatures and exceedingly long mission durations). The necessary work will include the development of key material testing capabilities that will enable material properties development from 40K – 400K for all properties needed to qualify space flight hardware (Tensile, Fatigue/Durability, etc.). The SMA wire properties will be input into a database and model, ultimately to disseminate material property design data (for 1 alloy only) as parameters for finite element modeling of a tire structure.

Andrea Marchica↗

Superconductivity devices: Commercial use of space

The work described in this report covers various aspects of the Rainbow solid-state actuator technology. It is presented in six parts dealing with materials, processing, fabrication, properties and associated phenomena. The Rainbow actuator technology is a relatively new materials development which had its inception in 1992. It consists of a new processing technology for preparing piezoelectric and electrostrictive ceramic materials. It involves a high temperature chemical reduction process which leads to an internal pre-stressing of the oxide wafer, thus the name Rainbow, an acronym for Reduced And INternally Biased Oxide Wafer. Ceramics fabricated by this method produce bending-mode actuator devices which possess several times more displacement and load bearing capacity than present-day benders (unimorphs, bimorphs). It is anticipated that these solid-state, electromechanical actuators which can be used in a number of applications in space such as cryopump motors, anti-vibration active structures, autoleveling platforms, telescope mirror correctors and autofocusing devices. When considering any of these applications, the key to the development of a successful device is the successful development of a ceramic material which can produce maximum displacement per volt input; hence, this initiative involving a solid-state means for achieving unusually high electromechanical displacement can be significant and far reaching. An additional benefit obtained from employing the piezoelectric effect in these actuator devices is the ability to also utilize them as sensors; and, indeed, they can be used as both motor (actuator) and generator (sensor) in multifunction devices.

Haertling, Gene↗

Development of Oxidation Protection Coatings for Gamma Titanium Aluminide Alloys

Metallic material systems play a key role in meeting the stringent weight and durability requirements for reusable launch vehicle (RLV) airframe hot structures. Gamma titanium aluminides (gamma-TiAl) have been identified as high-payoff materials for high-temperature applications. The low density and good elevated temperature mechanical properties of gamma-TiAl alloys make them attractive candidates for durable lightweight hot structure and thermal protection systems at temperatures as high as 871 C. However, oxidation significantly degrades gamma-TiAl alloys under the high-temperature service conditions associated with the RLV operating environment. This paper discusses ongoing efforts at NASA Langley Research Center to develop durable ultrathin coatings for protecting gamma-TiAl alloys from high-temperature oxidation environments. In addition to offering oxidation protection, these multifunctional coatings are being engineered to provide thermal control features to help minimize heat input into the hot structures. This paper describes the coating development effort and discusses the effects of long-term high-temperature exposures on the microstructure of coated and uncoated gamma-TiAl alloys. The alloy of primary consideration was the Plansee alloy gamma-Met, but limited studies of the newer alloy gamma-Met-PX were also included. The oxidation behavior of the uncoated materials was evaluated over the temperature range of 704 C to 871 C. Sol-gel-based coatings were applied to the gamma-TiAl samples by dipping and spraying, and the performance evaluated at 871 C. Results showed that the coatings improve the oxidation resistance, but that further development is necessary.

Wallace, T. A.↗

Extreme Spectroscopy: In situ nuclear materials behavior from optical data

In the nuclear industry, materials are regularly exposed to high temperature or/and irradiation and a better knowledge and understanding of their behavior under such extreme conditions is a key-point for improvements and further developments. Nowadays, Raman spectroscopy begins to be well known as a promising technique in the post mortem and remote characterization of nuclear materials exposed to extreme conditions. On this topic, at ANIMMA 2013 conference, we have presented some results about its implementation in the study of model or real nuclear fuel. However, the strength of Raman spectroscopy as in situ characterization tool is mainly its ability to be implemented remotely through optical fibers. Aware of this, implementation of other optical techniques can be considered in order to gain information not only on the structural dynamics of materials but also on the electronic charge carrier populations. In this paper, we propose to present our last advances in Raman characterization of nuclear materials and enlarge to the in situ use of complementary optical spectroscopies. Emphasis will be made on the information that can be gained to the behavior of the model fuel depleted UO{sub 2} under extreme conditions of high temperature and ionic irradiation: - In Situ Raman identification of the radiolysis alteration products of UO{sub 2} in contact with water under ionic irradiation. - In Situ Raman recording of the damaged dynamic of UO{sub 2} under inert atmosphere. - In Situ Raman and photo-luminescence study of virgin and damaged UO2 at high temperature. - In Situ study of electronic charge carriers' behavior in U{sub x}Th{sub 1-x}O{sub 2} solid solutions by mean of Iono- and Thermo- luminescence under and post- ionic irradiation. (authors)

36 MATERIALS SCIENCE↗

Fibre optic extensometer for high radiation and high temperature nuclear applications

In the framework of the Joint Instrumentation Laboratory (LCI), gathering resources from SCK-CEN (Belgium) and CEA (France), we are developing an optical sensor in order to accurately measure radiation-induced elongation of material placed in the core of a Material Testing Reactor (MTR). This extensometer displays common advantages of Fibre Optic (FO) sensors: high resolution, easy remote sensing and multiplexing, and also compact size which is of particular interest for in pile experiments with little room available. In addition, light weight reduces gamma heating hence limiting the thermal effect. In accordance with the specifications, the sensor has preferably two fixing points defining a gauge length of 10 to 15 mm (as a minimum). The diameter is less than 2 mm. Intense gamma and neutron irradiation as well as high temperatures are the most difficult environment conditions to withstand. Reactor radiation produces huge losses in common optical fibre. The losses can be limited by selecting the fibres (mainly with high purity silica core), the wavelength range (800-1200 nm), and a measurement based on interferometry (largely insensitive to losses in the fibre thanks to the wavelength encoding of the useful signal). Heavy neutron - mainly - and gamma flux such as in MTR, also produce compaction of silica, resulting in a significant drift and preventing the use of commercial FO sensors in such environment. Knowing this issue we revised the basic scheme of Extrinsic Fabry Perot Interferometer (EFPI) in order to limit the effects of compaction. A first sensor prototype fixed on a stainless steel support was tested in the Smirnof test facility in the BR2 MTR in Mol (Belgium). The support was subject to a constant mechanical and thermal stress, and his dimensions were not supposed to vary. This test showed a very low drift of the revisited EPFI design under high irradiation field in comparison with a commercial EFPI. This result has to be confirmed with second generation sensors with an increased robustness. The other difficulty to face is high temperature. Fibre optics with metal coating allows safe operation under temperatures up to 400 deg.C and even higher. But differential dilatation between silica and typical metallic material produces differential elongation in the range of 0.5 10{sup -2} i.e 5000{mu}{epsilon} for an increase in temperature of 300-400 deg. C. Such large elongation has to be considered carefully in the sensor design and its fixing on the sample. Laboratory tests with a second generation of adapted sensors demonstrated their good accuracy: less than 1% error over a displacement of 100{mu}m (< 1 {mu}m error), at room temperature. The displacement measured can reach 250{mu}m (20000{mu}{epsilon} with a gauge length of 12.5 mm) or more. We are currently implementing metal-coated fibre and we are preparing the next in pile irradiation on the BR2 reactor scheduled for mid-2011. Other applications of the sensor can be considered. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Stable platinum nanoparticles on specific MgAl 2 O 4 spinel facets at high temperatures in oxidizing atmospheres

The development of thermally stable, nanometer-sized precious metal-based catalysts remains a daunting challenge. Such materials, especially those based on the use of costly platinum metal, are essential and, to date, non-replaceable for a large number of industrially important catalytic processes. Here we report a well-defined cuboctahedral MgAl 2 O 4 spinel support material that is capable of stabilizing platinum particles in the range of 1–3 nm on its relatively abundant {111} facets during extremely severe aging at 800 °C in air for 1 week. The aged catalysts retain platinum dispersions of 15.9% with catalytic activities for methanol oxidation being ~80% of that of fresh ones, whereas a conventional Pt/γ-Al 2 O 3 catalyst is severely sintered and nearly inactive. Finally, we reveal the origin of the markedly superior ability of spinel {111} facets, resulting from strong interactions between spinel surface oxygens and epitaxial platinum {111} facets, inspiring the rational design of anti-sintering supported platinum group catalysts.

anti-sintering↗

Material development for thermionic fuel-cladding systems.

Fuel-cladding systems capable of maintaining stable dimension and electron emission characteristics at high temperatures under irradiation are essential to the successful operation of nuclear thermionic power sources. Studies regarding two possible systems, involving tungsten clad uranium carbide and uranium oxide, are considered. Out-of-pile developmental efforts and the in-pile evaluation results for the two systems are described. Prototypical emitters have been tested for 8000 hours for the carbide system and 4000 hours for the oxide system.

Yang, L.↗

Designing Magnetic Properties in CrSBr through Hydrostatic Pressure and Ligand Substitution

Abstract Magnetic van der Waals (vdW) materials are a promising platform for producing atomically thin spintronic and optoelectronic devices. The A‐type antiferromagnet CrSBr has emerged as a particularly exciting material due to its high magnetic ordering temperature, semiconducting electrical properties, and enhanced chemical stability compared to other vdW magnets. Exploring mechanisms to tune its magnetic properties will facilitate the development of nanoscale devices based on vdW materials with designer magnetic properties. Here it is investigated how the magnetic properties of CrSBr change under pressure and ligand substitution. Pressure compresses the unit cell, increasing the interlayer exchange energy while lowering the Néel temperature. Ligand substitution, realized synthetically through Cl alloying, anisotropically compresses the unit cell and suppresses the Cr‐halogen covalency, reducing the magnetocrystalline anisotropy energy and decreasing the Néel temperature. A detailed structural analysis combined with first‐principles calculations reveals that alterations in the magnetic properties are intricately related to changes in direct Cr–Cr exchange interactions and the Cr–anion superexchange pathways. Further, it is demonstrated that Cl alloying enables chemical tuning of the interlayer coupling from antiferromagnetic to ferromagnetic, which is unique among known two‐dimensional magnets.

Telford, Evan J.↗

Development of Austenitic ODS Strengthened Alloys for Very High Temperature Applications

This “Blue Sky” project was directed at exploring the opportunities that would be gained by developing Oxide Dispersion Strengthened (ODS) alloys based on the Fe-Cr-Ni austenitic alloy system. A great deal of research effort has been directed toward ferritic and ferritic/martensitic ODS alloys which has resulted in reasonable advances in alloy properties. Similar gains should be possible with austenitic alloy which would also take advantage of other superior properties of that alloy system. The research effort was aimed at the developing an in-depth understanding of the microstructural-level strengthening effects of ODS particles in austentic alloys. This was accomplished on a variety of alloy compositions with the main focus on 304SS and 316SS compositions. A further goal was to develop an understanding other the role of ODS particles on crack propagation and creep performance. Since these later two properties require bulk alloy material which was not available, this work was carried out on promising austentic alloy systems which could later be enhanced with ODS strengthening. The research relied on a large variety of micro-analytical techniques, many of which were available through various scientific user facilities. Access to these facilities throughout the course of this work was instrumental in gathering complimentary data from various analysis techniques to form a well-rounded picture of the processes which control austenitic ODS alloy performance. Micromechanical testing of the austenitic ODS alloys confirmed their highly superior mechanical properties at elevated temperature from the enhanced strengthening effects. The study analyzed the microstructural mechanisms that provide this enhanced high temperature performance. The findings confirm that the smallest size ODS particles provide the most potent strengthening component. Larger particles and other thermally- driven precipitate structures were less effective contributors and, in some cases, limited overall properties. With this understanding, the major materials development challenge is to provide a high uniformly distributed population of very fine ODS particles to be able to realize the full promise of dispersion strengthening. This should be a major goal of future work. This program had the further goal to develop graduate student researcher with the experience and capabilities to move this field forward. The support in this program was used for graduate student support and for research expenses; none of the program funds directly supported the faculty in the program. In this sense, the program was successful in supporting several very promising graduate researchers. Four of the graduate students supported here will complete their PhDs in 2015.

36 MATERIALS SCIENCE↗

Helium Irradiation Tolerance of Ti{sub 3}SiC{sub 2} MAX Phase Material

Fusion energy, which is the most possible way to completely solve human's energy problems, is facing with challenges on both theoretical and technological. In which the properties of the materials used in the fusion reactor are the most serious problems. What's more, several next generation nuclear power plant designs require materials to perform at temperature up to 1000+ deg. C in extreme conditions including high flux 14 MeV fast neutrons, large amount of evolutionary helium and hydrogen atoms and plasma facing environments. Development of new materials can normally work at extreme environments is one of the most important problems to solve before large scale application. The ternary layered MAX-phase materials combining the merits of both metals and ceramics, which could have high radiation damage tolerance and novel damage self-healing abilities due to their special structures, are considered as the candidate structural materials for next generation fusion reactor.

36 MATERIALS SCIENCE↗

NASA space applications of high-temperature superconductors

The application of superconducting technology in space has been limited by the requirement of cooling to near liquid helium temperatures. The only means of attaining these temperatures has been with cryogenic fluids which severely limits mission lifetime. The development of materials with superconducting transition temperatures (T sub c) above 77 K has made superconducting technology more attractive and feasible for employment in aerospace systems. Potential applications of high-temperature superconducting technology in cryocoolers and remote sensing, communications, and power systems are discussed.

Vernon O. Heinen↗

AGC-4 Graphite Preirradiation Data Analysis Report

The Advanced Reactor Technology (ART) Graphite R&D program is conducting an extensive graphite irradiation program to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs. Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of new HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each major graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change—where the graphite physically changes (i.e., first shrinking and then expanding with increasing neutron dose). This disparity in material volume change can produce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphite is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment. The AGC experiment is currently underway to determine the in service behavior of these new graphite grades for HTR and molten salt reactor designs. This experiment will examine properties and behavior of nuclear grade graphite over a large spectrum of temperatures, irradiation fluencies and applied stress levels that are expected to cause irradiation creep strains within a HTR graphite component. Irradiation data are provided through the AGC test series, which is comprised of six planned capsules irradiated in the Advanced Test Reactor (ATR) in a large flux trap located at Idaho National Laboratory (INL). Each irradiation capsule consists of over 400 graphite specimens that are characterized before and after irradiation to determine the irradiation induced material properties changes and life-limiting irradiation creep rate for each graphite grade.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The Path towards Plasma Facing Components: A Review of State-of-the-art in W-Based Refractory High-Entropy Alloys

Developing advanced materials for plasma-facing components (PFCs) in fusion reactors is a crucial aspect for achieving sustained energy production. Tungsten (W) - based refractory high-entropy alloys (RHEAs) have emerged as promising candidates due to their superior radiation tolerance and high-temperature strength. This review paper will focus on recent advancements in W-based RHEA research, particularly emphasizing the key role of modelling using machine learning (ML) in the stage of discovery by predicting properties for each composition and expediting the identification of optimal RHEA compositions with desired properties. Additionally, the application of additive manufacturing (AM) techniques for fabricating W-based RHEAs is explored, highlighting their advantages for rapid prototyping and multi-compositional sample production in a high throughput manner. The review critically evaluates the current understanding of mechanical properties relevant to PFC applications, including high-temperature strength and ductility. Furthermore, the radiation tolerance of W-based RHEAs under irradiated conditions is discussed. Finally, the validity of current AM-manufactured W-based RHEAs as PFC materials is assessed, and key challenges and opportunities for future research are identified. This review aims to provide a comprehensive overview of W-based RHEAs for fusion applications and their potential to guide the development and validation of advanced refractory high entropy alloys.

Hatler, Caleb [University of Wisconsin-Madison]↗

Low Temperature Regolith Bricks for In-Situ Structural Material

Current technology for producing in-situ structural materials on future missions to Mars or the moon relies heavily on energy-intensive sintering processes to produce solid bricks from regolith. This process requires heating the material up to temperatures in excess of 1000 C and results in solid regolith pieces with compressive strengths in the range of 14000 to 28000 psi, but are heavily dependent on the porosity of the final material and are brittle. This method is currently preferred over a low temperature cementation process to prevent consumption of precious water and other non-renewable materials. A high strength structural material with low energy requirements is still needed for future colonization of other planets. To fulfill these requirements, a nano-functionalization process has been developed to produce structural bricks from regolith simulant and shows promising mechanical strength results. Functionalization of granular silicate particles into alkoxides using a simple low temperature chemical process produces a high surface area zeolite particles that are held together via inter-particle oxygen bonding. Addition of water in the resulting zeolite particles produces a sol-gel reaction called "inorganic polymerization" which gives a strong solid material after a curing process at 60 C. The aqueous solution by-product of the reaction is currently being investigated for its reusability; an essential component of any ISRU technology. For this study, two batches of regolith bricks are synthesized from JSC-1A; the first batch from fresh solvents and chemicals, the second batch made from the water solution by-product of the first batch. This is done to determine the feasibility of recycling necessary components of the synthesis process, mainly water. Characterization including BET surface area, SEM, and EDS has been done on the regolith bricks as well as the constituent particles,. The specific surface area of 17.53 sq m/g (average) of the granular regolith material was obtained from nitrogen adsorption isotherm measurement. The size, shape and textures of regolith from SEM shows that the particles are 25-50 micrometers in size and mostly irregular in shape (Figure 1a). The elemental composition of regolith was identified from EDS analysis showed the presence of Si, Al, Fe, Na, Mg, Ca, Ti, O and C (see figure 1b). Each set of cylindrical brick samples were prepared by low energy process, and cured for 21 and 28 days, respectively to compare their compressive strength. Figure 1c, and d shows the JSC-1A brick and the compressive strength measurements. The results from the 21 day cured bricks (2 bricks) have been done and yielded an aver-age strength of 3050 psi, considerably higher than Portland cement mortars (Type IV and V). This promising technology provides the benefits of construction material similar to concrete, with a low complexity, low energy synthesis process and the likelihood of complete reusability of precious resources. Compressive strength using this method can be improved by increasing the surface area of the particles, using bi-modal particle size distribution, and adding certain additives to increase inter-particle forces.

ISRU↗

Manufacturability of U{sub 3}Si{sub 2} and its high temperature oxidation behaviour

High density fuels, such as uranium silicide (U{sub 3}Si{sub 2}), are being considered alongside improved cladding materials to develop new fuels which are both economic to manufacture and demonstrate enhanced accident tolerance. Through a collaborative programme led by Westinghouse and involving national laboratories in the US and UK, efforts are being made to establish a viable route by which U{sub 3}Si{sub 2} can be manufactured and to test its suitability as a fuel. In the UK work has focused on routes to convert UF{sub 6} and UF{sub 4} to U{sub 3}Si{sub 2} directly, without needing to reduce the uranium to its metallic state. Experimental observations of reactions between UF{sub 4} and Si indicate the formation of a USi{sub 2-x} phase at a temperature of 1000 C. degrees, consistent with thermodynamic assessments. Work on the construction of a rig which will pass UF{sub 6} heated to 1000 C. degrees through a bed of finely divided Si powder, is also reported. At Westinghouse, U{sub 3}Si{sub 2} fabricated by the arc-melting process has been subjected to high temperature steam and synthetic air tests and the reaction rate measured as a function of temperature using differential scanning calorimetry. The rates are compared to UO{sub 2} and UN (Uranium mononitride). At Idaho National Laboratory irradiation testing of U{sub 3}Si{sub 2} fuel is in progress. Fuel specimens were prepared by conventional arc melting process to formulate the U{sub 3}Si{sub 2} followed by conventional powder metallurgy processes to create pellets. These irradiation tests are intended to help resolve questions concerning the swelling of U{sub 3}Si{sub 2} under conditions representative of LWR fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Temperature and dose effects on dislocation loops in self-ion irradiated high-purity iron

Body-centered cubic (BCC) Fe-based alloys are promising candidate materials for advanced nuclear reactors. However, a detailed understanding of irradiation induced dislocation loop microstructure development remains unresolved. It is a widespread belief that 〈001〉 loops become increasingly favorable over ½〈111〉 loops as irradiation temperature rises above ∼300 C. Unfortunately, the temperature effects on 〈001〉 loop have been primarily examined in in-situ irradiation on TEM thin foils but poorly explored on bulk Fe due to exceedingly limited experimental studies on bulk specimens, raising concerns about the potential influence of TEM thin foil artifacts on observed results. Here, in this study, we conducted experiments on ultra-high purity BCC Fe specimens irradiated with 6.7–8 MeV Fe ions over a wide temperature range on bulk samples. We investigated the effects of temperature (T irr = 250–500 °C), dose rate (10⁻⁵ to 10⁻³ dpa/s), and dose (0.35 to 3.5 dpa) on the formation and evolution of 〈001〉 and ½〈111〉 loops as well as cavity (void) formation. Post-irradiation Burgers vector analysis via g•b method on dislocation segments and loops revealed that 〈001〉 loop fraction does not show a monotonic positive correlation with irradiation temperature. Combined with previous and current theoretical as well as experimental findings, we explore the temperature effects on all existing models of 〈001〉 loop formation. We conclude that the prevailing reports regarding the dominance of 〈001〉 loops in Fe at elevated temperatures are mainly attributable to the loss of glissile ½〈111〉 clusters in TEM thin foil experiments.

36 MATERIALS SCIENCE↗

Performance Modeling of Materials-Based Hydrogen Storage Systems for Automotive Applications

One of the challenges of fuel cell vehicles is storage of the hydrogen fuel. In an effort to reduce the temperature and pressure limitations of storing neat hydrogen either as a high pressure gas or cryogenic liquid, material-based hydrogen storage is being developed. Hydrogen stored either as an adsorbent or as a chemical compound not only provides more reasonable pressures and temperatures but also improves the storage density of the hydrogen. Modeling tools have been developed as part of the U.S. Department of Energy's Hydrogen Storage Engineering Center of Excellence to assess the performance of these materials in light duty vehicle applications. The purpose of these modeling tools is to evaluate newly developed hydrogen storage materials and ensure that they meet customer expectations for driving range and performance before further investment. Recent advances in these modeling tools have been made to allow direct material property inputs. Such material properties as hydrogen capacity, isotherms and other thermodynamic and kinetic properties can reasonably be measured by materials researchers. Using these properties as inputs, the models can then provide estimates of the mass and volume of the storage system and the driving range and fuel efficiency of the vehicle during transient operation with four different drive cycles. The system sizing estimator and the framework model used to perform this analysis will be described and results presented for adsorbents and chemical based storage materials.

25 ENERGY STORAGE↗