Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear alloys”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Bridging microscale to macroscale mechanical property measurements and predication of performance limitation for FeCrAl alloys under extreme reactor applications

Microscale mechanical testing has greatly benefited nuclear materials studies in at least two aspects: one is its feasibility to integrate with SEM and TEM microscopes for in situ atomic scale or microscale structural characterization to reveal fundamental details, and the other is its significance in development of accelerator-based ion irradiation technique as a surrogate method to simulate neutron damage. Ion irradiation is able to reach damage creation at a level at least three orders of magnitudes higher than test reactors. However, limited ion penetration depths, which are about a few microns for MeVs heavy ions and 10s microns for MeV light ions, make microscale mechanical tests a necessity. But, there is a great challenge to bridge microscale tests to macroscale tests because a bulk specimen of irradiated nuclear materials for high dose applications cannot be obtained in laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Microstructural evolution in doped high entropy alloys NiCoFeCr-3X (X=Pd/Al/Cu) under irradiation

Commonly studied equatomic single-phase FCC high entropy alloys based on 3d transition metals like NiCoFeCr do not provide adequate strength and radiation resistance at high doses for nuclear structural applications. In the current study, the major alloying effects like lattice distortion, ordering and clustering tendencies were investigated by adding low concentration of Pd, Al, or Cu respectively to study the doping effects on the ion irradiation response of NiCoFeCr alloy. The alloys were irradiated with 3 MeV Ni 2+ ions at 500 °C to a fluence of 1 × 10 17 /cm 2 at a beam flux of approximately 2.8 × 10 12 ions/cm 2 /s. The microstructural evolution upon irradiation i.e., formation of dislocation networks, radiation induced segregation and precipitation, and void formation were studied in detail. Further, post-irradiation characterization results showed that a Pd addition leads to a high void nucleation rate but controlled void growth, which may be attributed to increased lattice distortion. In Al added HEA, our microstructural analysis indicates that radiation induced ordered L1 2 precipitates do not affect void swelling significantly. Cu addition led to Cu precipitation that drastically suppressed dislocation density and void swelling of the alloy. Additionally, a model was developed to qualitatively describe the trend in void swelling of typical FCC alloys under ion irradiation. This model was able to qualitatively explain the suppression and reappearance of void swelling in ion irradiated alloys that generally occurs near the region with peak implanted ion concentration.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 20326

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering scale. Veolia Nuclear Solutions designed, installed, and commissioned in September 2018, at Perma-Fix Northwest in Richland Washington, a 10-metric ton full-scale GeoMelt unit (GeoMelt{sup R} Richland) for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt{sup R} Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development

A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States’ high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo fuel and the data produced includes bending strength and Young’s modulus. Although the material behaves in a brittle manner due to the developed porosity, a general trend of strength and modulus reduction are found as fission density increases. The data produced is evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effect of Li metal addition on corrosion control of Hastelloy N and stainless steel 316H in molten LiF-NaF-KF

Molten fluoride salts are used in various energy harvesting applications such as solar collectors and molten salt nuclear reactors (MSRs). Thermodynamically driven selective dissolution of alloying elements due to impurities present within the fluoride salts have been identified to be the main corrosion mechanism for structural alloys in molten fluoride salts. Impurities in these salts, during salt preparation as well as during operation, can be difficult to control. One way to prevent the selective dissolution of active alloying elements is to make the molten salt reducing by either purifying it to remove impurities or to add some reactive metal that will react with the impurities and make the salt reducing. Here this study is focused on the corrosion behaviour of Hastelloy N and 316H stainless steel in molten LiF-NaF-KF eutectic mixture (FLiNaK) at 700°C and how this behaviour changes with the addition of Li metal as a reducing impurity. Our results indicate that the active metal addition like Li can be very effective in controlling the redox behaviour of salt, which in-turn can mitigate corrosion. Even the addition of small quantities of Li metal can mitigate the selective dissolution of active alloying elements for the tested structural materials in FLiNaK. Our results also suggest that the presence of excess amounts of Li, above the minimum concentration required to react with the impurities in FLiNaK, does not have any detrimental effect in terms of formation of new intermetallic phases at the surface of the selected alloys under tested conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Silicon carbide reinforced zirconium based cladding

A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Evolution of Microstructure and Surface Characteristics of FeCrAl alloys when Subjected to Flow Boiling Testing

FeCrAl alloys are candidate materials for manufacturing accident-tolerant fuel (ATF) cladding intended for light water reactors to increase fuel reliability and safety during design-basis and beyond-design-basis accident scenarios. The evolution of the materials' surface characteristics, microstructure, and mechanical properties when exposed to the Critical Heat Flux (CHF) in flow boiling testing is crucial for safety analysis while providing insights into their thermal-hydraulic performance in nuclear reactors. After CHF, the surface chemistry of two FeCrAl alloys, APMT and C26M, was studied to understand their evolution at the early stage of high-temperature excursions in short time periods. The results indicated a thin layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the layer, as indicated by X-ray photoelectron spectroscopy (XPS) and depth profiling. The cross-sections prepared by focused ion beam (FIB) revealed the growth of an oxide layer, in the range of 90-180 nm thick, on the alloys' surfaces. The evolution of the materials' surface chemistry also led to a noticeable post CHF excursion increase in their wettability, with a slight increase in roughness. Further, the investigation of the materials' mechanical properties indicated a modest increase in hardness by 10-15% as well as an increase in their yield strength, as evidenced by the microindentation and ring compression tests conducted before and after CHF testing. Scanning electron Microscopy (SEM) and X-ray diffraction (XRD) were used to investigate microstructural features of the materials and their changes after CHF treatment.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Crystallographic Orientation Dependence of Mechanical Responses of FeCrAl Micropillars

Iron-chromium-aluminum (FeCrAl) alloys are used in automobile exhaust gas purifying systems and nuclear reactors due to its superior high-temperature oxidation and excellent corrosion resistance. Single-phase FeCrAl alloys with a body centered cubic structure plastically deform through dislocation slips at room temperature. Here, we investigated the orientation dependence of mechanical responses of FeCrAl alloy through testing single-crystal and bi-crystal micropillars in a scanning electron microscopy at room temperature. Single-crystal micropillars were fabricated with specific orientations which favor the activity of single slip system or two slip systems or multiple slip systems. The strain hardening rate and flow strength increase with increasing the number of activated slip system in micropillars. Bi-crystal micropillars with respect to the continuity of slip systems across grain boundary were fabricated to study the effect of grain boundary on slip transmission. The high geometrical compatibility factor corresponds to a high flow strength and strain hardening rate. Experimental results provide insight into understanding mechanical response of FeCrAl alloy and developing the mechanisms-based constitutive laws for FeCrAl polycrystalline aggregates.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Assessment of Process Modeling Tools for Determining Variability in Additively Manufactured Parts

The Advanced Materials and Manufacturing Technologies (AMMT) program aims to accelerate the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy However, the unique aspects of additive manufacturing (AM) materials in terms of their processing history, microstructure, and properties, are a major barrier for qualification and certification of nuclear components. Much of this challenge may be attributed to component scale variations in microstructure and properties that are driven by local influences of process conditions and geometry on thermal history, melt pool dynamics, and corresponding microstructure evolution. Computational modeling tools may be helpful in this regard to aid in predicting and controlling this level of variability. The purpose of this report is to review the current state-of-the-art for process modeling with regards to metal AM. For this purpose, we consider specifically the case study of laser powder bed fusion (LPBF) processing of SS316, a family of alloys that are both commonly used in nuclear energy applications and suitable for AM processing. The report first introduces the necessary components of a process modeling workflow, followed by a review of the current status of each. At the end, application of these modeling tools to understanding variability in AM process given their current state are considered, and recommendations for future development are proposed

36 MATERIALS SCIENCE↗
DOE OSTI2021

Fundamental Studies of Tritium Formation and Diffusivity in Pure and Defective Zircaloy-4 Getters

Zirconium (Zr) alloys have excellent mechanical properties at high temperature and are resistant to corrosion in different environmental conditions. These alloys are used as tritium (3H or T) getters in nuclear reactors due to their low absorption cross section to thermal neutrons and excellent thermo-mechanical properties. Zr alloys have better mechanical and thermal properties than many other refractory alloys including stainless steel. After the irradiation of TPBAR by neutron flux, 3H produced during irradiation diffuses though the pellets and is captured by the getter. 3H chemically reacts with Zr metal to form metal hydride (ZrTx). Hydrides thus formed are brittle and adversely affect the mechanical and thermal properties of alloys. The lattice mismatch at the interface of alloy and metal hydrides also creates a stress that significantly reduces the alloy’s performance and operation life. Therefore, understanding the behavior of 3H and its species becomes significant as fuel burnup is increased, which leads to increase in hydrogen pickup and oxide formation.

Duan, Yuhua↗
DOE OSTI2022

IER 517: Molybdenum Critical Experiment Design [Slides]

In nuclear fuels, U-Mo metallic fuels are used in new space reactor designs and new research reactor fuels. MITR, MURR, and NBSR reactors plan on converting from different HEU fuels to a U-Mo HALEU fuel. In spend nuclear fuel, 95 Mo is one of the 15 main absorbing fission products in irradiated LWR fuel. This makes it important for criticality safety studies in transportation and reprocessing. In structural materials. Molybdenum is found commonly in alloys that make up the structural materials of nuclear reactors such as type 316 SS. Molybdenum helps to improve high temperature performance and corrosion resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Heavy ion irradiation response of an additively manufactured 316LN stainless steel

Additive manufacturing has become an appealing technique to fabricate three-dimensional metallic materials and components for nuclear reactors. However, response of additively manufactured alloys to high-dose heavy ion irradiations at elevated temperatures is still not well understood. Here, an additively manufactured 316LN austenitic stainless steel with high-density solidification cells was irradiated using 3.5 MeV Fe ion to a peak dose of 220 dpa at 450 °C. Microscopy studies show a lower Frank loop density and smaller size in the additively manufactured sample compared with its cold worked counterpart, and the cellular structures may largely suppress the formation of perfect loops and dislocation networks and reduce the magnitude of solute segregations comparing with high angle grain boundaries. In conclusion, the present work advances the understanding on the high-temperature irradiation response of additively manufactured steels for nuclear reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Report Summarizing the Mechanical Properties of a Large ODS Ferritic Alloy Ingot by Forging at High Temperatures for Future Mother Tube Production

Oxide dispersion strengthened (ODS) ferritic alloys are considered the benchmark fuel cladding and core structural material in advanced nuclear energy reactors that require high-temperature strength and creep properties and resistance to radiation damage. The ODS ferritic alloys including 14YWT are produced by mechanical alloying (MA), which is time-consuming and is associated with high manufacturing costs that are not beneficial to being used in advanced nuclear energy reactors. This report summarizes the accomplishments in FY24 in the INM program for producing an ODS ferritic alloy by high-deformation, high-temperature processing of reactive and ferritic alloy powders for advanced reactor fuel cladding applications. Following four hot forging experiments, it was concluded that several significant challenges were encountered that were difficult to overcome. These challenges were related to using high annealing temperatures for pressure assisted sintering processes to produce dense microstructures, but the high annealing temperatures result in long range diffusion of Ti through the bcc Fe lattice and react with the YIG particles that are distributed on the prior surfaces of the ferritic alloy powders. In addition, it was determined that the high deformations induced by forging at high temperatures were not very effective for fracturing the reactive YIG particles into smaller particles and distributing them into the interior of the ferritic alloy powders. Spark plasma sintering was attempted for shortening the time at high temperatures, which led to full densification but the microstructure characterization results still showed that Ti atoms diffused over long distances in the bcc Fe lattice and reacted with the YIG particles. Finally, a short 30 minute high intensity ball milling experiment was conducted on blended 14WT and YIG powders for inducing severe deformations that caused the initially spherical powders of 14WT decorated with the YIG particles on the surfaces to transform to flakes and fracturing of YIG particles into smaller particles that were incorporated into the bcc Fe lattice. Two forgings were performed with the high intensity milled powders: first annealing for 20 minutes at 850ºC followed by forging and second annealing for 20 minutes at 1,100ºC followed by forging. At 850ºC, Ti atoms are effectively immobile, thus allowing for densification albeit incomplete. The deformation by forging after annealing for 20 minutes at 1,100ºC favored dynamic recrystallization processes that led to nano-size grains with very high stored energy due to high dislocation density. The corresponding VH data showed significant hardening that correlated with estimated strengthening of ~1650 MPa. This hybrid processing approach combining high intensity ball milling of powder with annealing and forging showed the most promise for producing an oxide dispersing strengthened ferritic alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Influence of rolling reduction and annealing on recrystallization and grain structure in Ta-2.5W alloys

The microstructural evolution of wrought Tantalum - 2.5 wt% Tungsten (Ta-2.5W) alloys during thermomechanical processing is critical for optimizing their mechanical reliability in demanding applications such as aerospace, chemical processing, and nuclear technology. Despite the widespread use of Ta-W alloys, a comprehensive understanding of how rolling reduction, annealing temperature, and elemental inhomogeneity interact to determine recrystallization behavior and grain refinement remains incomplete. Here, in this study, we systematically investigate the effects of cold rolling and subsequent annealing on the microstructure of Ta-2.5W, with particular attention to grain orientation, stored energy, and elemental banding. Our results demonstrate that higher rolling reduction rates lower the onset and completion temperatures for recrystallization, resulting in finer and more homogeneous grain structures. Electron backscatter diffraction (EBSD) analysis reveals that grains with a 〈111〉 parallel to the plate normal possess higher stored energy and nucleate recrystallization more readily than grains with a 〈001〉 parallel to the plate normal. Elemental mapping shows that tungsten inhomogeneity leads to localized bands of accelerated recrystallization and hardness variation. These findings provide new insights into the mechanisms of microstructure refinement in Ta-2.5W alloys, offering guidance for tailoring processing routes to achieve superior performance in demanding engineering environments.

Annealing↗
DOE OSTI2020

Alloys for inert matrix fuel compositions, and methods of making the same

In one embodiment, an alloy includes: Zr; Fe; Cu; Ta in an amount from about 1 wt % to about 3 wt %; and one or more optional constituents selected from: Ti, Be, and Nb; and wherein the alloy comprises a ductile phase and a nanoprecipitate hard phase. According to another embodiment, a method of forming an inert matrix nuclear fuel includes: packing a hollow structure with fuel pellets and alloy precursor pellets; heating the fuel pellets and the alloy precursor pellets to at least a melting temperature of an alloy to be formed by melting the alloy precursor pellets; and solidifying the alloy into a matrix surrounding the fuel pellets. The alloy precursor pellets independently comprise: Zr; Fe; Cu; Ta present in an amount from about 1 to about 3 wt %; and one or more optional alloy constituents selected from: Ti, Be, and Nb.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Enhanced diffusion bonding of alloy 617 using electric field-assisted sintering

The development of compact heat exchangers (CHXs) has gained increasing interest in many industries owing to their high thermal efficiency and reduced size. Diffusion bonding (DB) is an advantageous technique for fabricating CHXs. Alloy 617 is a candidate for manufacturing CHXs for high-temperature advanced nuclear reactors due to its elevated-temperature properties. Previous endeavors in DB of Alloy 617 were conducted by hot pressing (HP), which reported precipitates at the diffusion-bond interface, limited grain boundary (GB) migration, and significantly reduced high-temperature mechanical properties. To overcome these challenges, this study investigated DB of Alloy 617 using electric field-assisted sintering (EFAS). Stacks composed of three sheets were bonded with EFAS using different temperatures, pressures, and hold times. DB using HP as the zero-current analog of EFAS was also performed for comparison. The result shows that Cr- and Mo-rich precipitates were formed at the interface of the hot-pressed samples. The electric current and temperature in EFAS play a significant role in precipitation and GB migration. The electric current coupled with correct temperatures can effectively prevent precipitate formation at the interface and achieve excellent GB migration. Nanoscale Al-rich oxide was formed at the interface of the samples made by both HP and EFAS, but grain boundaries can ignore the nanoscale Al-oxide and migrate across the interface. The temperature, pressure, and hold time also affected diffusion. The temperature is a prerequisite for a successful GB migration, and GB migration can be enhanced by increasing pressure and hold time.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Irradiation-induced shift in the thermodynamic stability of phases and the self-healing effect in transformative high entropy alloys

Here, this work investigated Fe 40 Mn 20 Cr 15 Co 20 Si 5 high entropy alloy (CS-HEA), which exhibits transformation induced plasticity (TRIP) from γ-fcc → ε-hcp, as a probable candidate for nuclear applications. CS-HEA is an extensively-explored, low stacking fault energy alloy with superior strength, ductility, fatigue resistance, and corrosion resistance. This study delved into the effect of irradiation on the shift in thermodynamic stability of the phases and thus radiation tolerance. The evolution of phases, lattice parameters, and transformation volume, V γ→ε , were evaluated from X-ray diffraction experiments along with the mechanical response from nanoindentation. The alloy exhibited a recently-proposed novel self-healing mechanism possible due to the TRIP effect to minimize irradiation damage by restraining the γ-fcc → ε-hcp transformation via thermal aid; this self-healing mechanism was confirmed by transmission electron microscopy. The results were corroborated by a negative change in V γ→ε and a low |V γ→ε |, which is an important criterion for recovery of parent γ-fcc phase. Thus, this alloy was deemed a good radiation-tolerant candidate for nuclear application.

36 MATERIALS SCIENCE↗