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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

The primary challenge preventing the deployment of reactor designs that leverage wrought FeCrAl as an advanced light-water reactor (LWR) cladding is irradiation hardening and embrittlement. Wrought FeCrAl alloys experience a loss of post-irradiation ductility and fracture toughness under low-temperature neutron irradiation (< 350°C)resulting from the combined effects of dislocation loop formation and the irradiation-enhanced precipitation of Cr-rich alpha-prime precipitates throughout the microstructure. Over the past decade, significant improvements in our understanding regarding the effect of Cr and Al content have been enabled through neutron irradiations, but even optimized wrought alloys such as C26M remain vulnerable to brittle failure during storage, transportation, and handling following irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

The primary challenge preventing the deployment of reactor designs that leverage wrought FeCrAl as an advanced light-water reactor (LWR) cladding is irradiation hardening and embrittlement. Wrought FeCrAl alloys experience a loss of post-irradiation ductility and fracture toughness under low-temperature neutron irradiation (< 350°C) resulting from the combined effects of dislocation loop formation and the irradiation-enhanced precipitation of Cr-rich alpha-prime precipitates throughout the microstructure. Over the past decade, significant improvements in our understanding regarding the effect of Cr and Al content have been enabled through neutron irradiations, but even optimized wrought alloys such as C26M remain vulnerable to brittle failure during storage, transportation, and handling following irradiation. This report summarizes multiple irradiation campaigns initiated over the past 7 years and provides recent insights into the effect of (1) minor alloying composition, (2) alloy processing strategy, (3) crystal structure, and (4) the use of advanced oxide dispersion strengthened (ODS) alloys. The first irradiation campaign irradiated body-centered cubic FeCrAl alloy C26M with and without elements such as Mo and Y. It also included face-centered cubic alumina-forming austenitic (AFA) alloys that have a steeper neutronic penalty but that have far superior high-temperature strength than wrought FeCrAl. Finally, this campaign compared conventional wrought C26M with the same material produced using modern powder metallurgical hotisostatic pressing (PM-HIP). The second irradiation campaign, initiated in 2018, irradiated wrought and FeCrAl-ODS materials up to 50 dpa to assess their irradiation resistance to high-doses. The results of this work indicate that wrought C26M suffers significant degradation following neutron irradiation at LWR-relevant temperatures regardless of Mo and Y content. The PM-HIP variant appears to be more resilient to irradiation-induced ductility loss in comparison with wrought variants but still loses almost all ductility by 8 dpa. Notably, the AFAs retained at least 5% total elongation after irradiation at the same dose and temperature condition, although additional deterioration may be expected at higher fluence levels. FeCrAl-ODS materials show the greatest promise as a transformative longer-term accident-tolerant fuel (LT-ATF) cladding material. The extruded 106ZY10C alloy retained 10% total elongation after 16 dpa irradiation and retained 8% total elongation after 50 dpa. Additional challenges with FeCrAl-ODS alloys also remain, including optimization of end cap joining methodologies, enhancement of fracture toughness, scaling production to prove economic viability, and optimizing post-pilger heat treatments to maximize ductility and irradiation resistance. This series of irradiations demonstrates the efficacy of the LWR testbed available within the United States for the rapid irradiation and down-selection of LT-ATF candidates. As the country rapidly accelerates its timeline for the deployment of advanced reactor concepts, the effective utilization of the High Flux Isotope Reactor for separate-effects style irradiations should continue to be prioritized to answer the final questions pertaining to LT-ATF candidates necessary for the deployment of advanced boiling water reactors and small modular reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Cluster Dynamics Simulations of Intra-Granular Fission Gas Bubble Size and Pressure Evolution in UO 2

Fission gases such as xenon (Xe) play a critical role in determining the behavior and response of nuclear fuel. Given that Xe has little solubility in UO 2 , it accumulates and forms bubbles, which significantly impact fuel performance. Intra- and inter-granular bubble nucleation and growth can lead to fuel swelling, and once bubbles interconnect at grain boundaries, fission gas can be released into the plenum. At low temperatures, limited uranium vacancy mobility can restrict swelling, therefore causing the bubbles to become highly pressurized. Consequently, this can induce micro-cracking, promote fission gas release (increasing the likelihood of cladding failure), and even lead to fuel pulverization under accident conditions such as a loss of coolant accident. As bubble evolution is strongly influenced by local temperature and fission rate, markedly different behavior occurs across the radial profile of the fuel pellet. Capturing the mechanisms that underpin bubble evolution is therefore important to predict these behaviors in the fuel. Previous models describing important mechanisms informed by lower length scale simulations have been developed under the NEAMS program. These can describe the evolution of a single bubble type (i.e., single value for radius and pressure) at each position in the pellet, for instance using the Centipede cluster dynamic code. However, in reality, a full distribution in bubble sizes and pressures exists within the microstructure at a given position in the pellet. To address this the cluster dynamics code Xolotl, which can predict Xe and vacancy phase space (i.e., bubble distributions) for intra-granular bubbles, has been used before. Prior work benchmarked the Xolotl code against the Centipede cluster dynamics code to ensure compatibility and to verify that mobile defect properties are adequately transferred between the two codes, along with some physics improvements. In this work, we go further by introducing a physics-based set of improvements that will allow us to accurately predict bubble size distributions and internal bubble pressures under representative UO 2 irradiation conditions. The improvements include (i) coupling bubble-defect reaction energies to a virial equation of state (EOS), (ii) including a bubble surface tension contribution, (iii) incorporating radiation-induced re-solution of Xe and vacancies, (iv) enabling pressure-driven dislocation loop punching through an effective emission of interstitial clusters informed by interstitial loop energetics, (v) accounting for radiation induced athermal diffusion of Xe, and (vi) implementing a Booth-type grain boundary sink representation for all mobile defects and defect clusters. After these modifications, we observe good agreement of Xolotl fission gas bubble size and concentration predictions with legacy experimental measurements. Additionally, it allows the distribution of Xe bubble pressures and radius to also be predicted and compared to data produced through the Advanced Fuels Campaign (AFC) program. Here, we have done this by running simulations under conditions similar to the AFC post-irradiation examination (PIE) samples irradiated at North Anna 2 light water reactor (LWR). Our results shows excellent agreement with these experimental measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Towards Ab-Initio Simulations of Crystalline Defects at the Exascale Using Spectral Quadrature Density Functional Theory

Defects in crystalline solids play a crucial role in determining properties of materials at the nano, meso- and macroscales, such as the coalescence of vacancies at the nanoscale to form voids and prismatic dislocation loops or diffusion and segregation of solutes to nucleate precipitates, phase transitions in magnetic materials via disorder and doping. First principles Density Functional Theory (DFT) simulations can provide a detailed understanding of these phenomena. However, the number of atoms needed to correctly simulate these systems is often beyond the reach of many widely used DFT codes. The aim of this article is to discuss recent advances in first principles modeling of crystal defects using the spectral quadrature method. The spectral quadrature method is linear scaling with respect to the number of atoms, permits spatial coarse-graining, and is capable of simulating non-periodic systems embedded in a bulk environment, which allows the application of appropriate boundary conditions for simulations of crystalline defects. In this article, we discuss the state-of-the-art in ab-initio modeling of large metallic systems of the order of several thousand atoms that are suitable for utilizing exascale computing resourses.

36 MATERIALS SCIENCE↗
NASA NTRS1971

Heterogeneous nucleation of gamma prime in Al-Ag and Al-Ag /Cd or Cu/ alloys.

Results of transmission electron microscopy studies of heterogeneous nucleation of the metastable gamma prime phase were made in binary and ternary alloys of Al-5 at. % Ag, Al-4.8 at. % Ag-0.15 at. % Cd, and Al-4.8 at. % Ag-0.2 at. % Cu. Heterogeneous nucleation of gamma prime was found to be associated with four types of dislocations, regardless of the quenching treatment or alloy additions which were used. Nucleation was found to occur on: (a) jogged dislocations; (b) helical dislocations; (c) dislocation loops; (d) heterogeneous climb sources. Only one nucleation mechanism was found to be unique to the ternary alloys: that of nucleation on heterogeneous climb sources.

Passoja, D. E.↗
NASA NTRS1982

Microstructural effects on the room and elevated temperature low cycle fatigue behavior of Waspaloy

Longitudinal specimens of Waspaloy containing either coarse grains with small gamma or fine grains with large gamma were tested in air at a frequency of 0.33 Hz or 0.50 Hz. The coarse grained structures exhibited planar slip on (III) planes and precipitate shearing at all temperatures. Cracks initiated by a Stage 1 mechanism and propagated by a striation forming mechanism. At 700 C and 800 C, cleavage and intergranular cracking were observed. Testing at 500 C, 700 C, and 800 C caused precipitation of grain boundary carbides. At 700 C, carbides precipitated on slip bands. The fine grained structures exhibited planar slip on (111) planes. Dislocations looped the large gamma precipitates. This structure led to stress saturation and propagation was observed. Increasing temperatures resulted in increased specimen oxidation for both heat treatments. Slip band and grain boundary oxidation were observed. At 800 C, oxidized grain boundaries were cracked by intersecting slip bands which resulted in intergranular failure. The fine specimens had crack initiation later in the fatigue life, but with more rapid propagation crack propagation.

Lerch, B. A.↗
NASA NTRS1988

Elevated temperature slow plastic deformation of NiAl/TiB2 particulate composites

The 'XD' process for production of discontinuously-reinforced metal-matrix composites has been used to enhance the high-temperature strength of NiAl-TiB2 composites with particulate densities of up to 30 vol pct. SEM, TEM, and optical characterizations of the resulting microstructures showed the average TiB2 particle size to be about 1 micron, while the average grain of the NiAl matrix was of the order of 10 microns. Elevated temperature compression tests conducted at 1200 and 1300 K indicated flow strengths to increase with TiB2 content, so that the 20 vol pct TiB2-reinforced composite was three times stronger than the unreinforced NiAl; this is ascribed to the very high density of microstructural tangled dislocations, loops, and subgrain boundaries connecting the particles.

Whittenberger, J. Daniel↗
NASA NTRS1959

Factors Affecting the Plasticity of Sodium Chloride, Lithium Fluoride, and Magnesium Oxide Single Crystals

A study was made of the relative magnitude of the effects of various factors on the ductility of single crystals of sodium chloride (NaCl), lithium fluoride (LiF), and magnesium oxide (MgO). Specimen treatments included water-polishing, varying cleavage rate, annealing, quenching, X-irradiation, surface coating, aging, and combinations of some of these treatments. The mechanical behavior of the crystals was studied in flexure and in compression, the latter study being performed at both constant strain rate and constant load. Etch-pit studies were carried out to provide some pertinent information on the results of pretreatment on the dislocation concentration and distribution in the vicinity of the surface. The load deformation curves for these ionic single crystals show an initial region of very low slope which proved to be due to anelastic deformation. The extent of initial anelastic deformation is modified by specimen pretreatment in a way that suggests that this deformation is the result of expansion of cleaved-in dislocation loops, which can contract on the removal of the stress. The effects of the various pretreatments on the load and deflection at fracture are in accord with the prediction one might make with regard to their effect on the nucleation of fatal surface cracks. For NaCl, increases in ductility are always accompanied by increases in strength. The creep constants for NaCl are a function of treatments which affect the bulk structure but are not a function of treatments which only affect the surface.

Stearns, Carl A.↗
DOE OSTI2021

Probing Microstructure-induced Swelling & Thermal Property Changes in Ion-Irradiated Oxide Fuels using Laser-generated Surface Acoustic Waves

Radiation of fuel rods with neutrons during a fission reaction in a nuclear reactor is known to affect the material microstructure of the fuel as a result of fission fragment damage, and lead to the formation of atomic-level defects such as voids, dislocation loops, and lattice swelling from fission-gas release. A consequence of the radiation-induced damage in nuclear fuels is the drastic alteration of material properties, in particular, the reduction of the thermal conductivity – a key parameter that governs the transport of thermal energy released from fissile fuel atoms to the surrounding coolant. Swelling can induce high stresses and ultimate failure of the cladding. A fundamental understanding of the role of radiation-induced damage on the thermal transport and mechanical properties of nuclear fuels is therefore critical for efficient, reliable and safe operation of a nuclear power plant. As opposed to conventional post-irradiation examination techniques that are often time-consuming and require significant sample preparation, laser-based characterization methods have emerged as promising non-contact, non-destructive tools to measure the evolution of microstructure-induced material property changes. Moreover, laser methods offer the promise of in-situ characterization while the material is being irradiated. This project aims to develop an improved understanding of the impact of radiation-induced changes to material microstructure on the thermal and elastic properties of nuclear materials. A laser ultrasonic technique, known as the transient grating (TG) spectroscopy method, is used to simultaneously measure thermal diffusivity and elastic properties in unirradiated and ion-irradiated oxide nuclear fuel samples. The results will complement on-going investigations on electron- and phonon-mediated thermal transport in nuclear materials and will provide foundational work for incorporating the influence of defects in fuel performance codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Impact of Irradiation on Microstructure and Mechanical Properties of Materials Produced by Advanced Manufacturing

Advanced non-light water reactor designs, known as Generation IV (Gen IV) reactors, typically operate at higher temperatures and under more extreme radiation conditions than conventional light water reactors. A critical aspect of the successful deployment and advancement of Gen IV reactor designs is the selection of appropriate structural materials for specific applications, which necessitates the timely development of new materials and manufacturing processes. Advanced manufacturing (AM) offers numerous opportunities for innovative designs, enabling the production of high-performance components with potentially shorter development cycles compared to traditional manufacturing methods. However, a significant challenge in deploying AM technologies in the nuclear energy sector is the current lack of data on the irradiation performance of AM-produced components. This presentation will discuss neutron and ion irradiation results of AM materials, including stainless steel 316L, Grade 91, SA508, Inconel 718, and Inconel 625 materials. The materials were manufactured by AM, such as Powder Metallurgy Hot Isostatic Pressing (PM HIP), Laser Powder Bed Fusion (LPBF) and Directed Energy Deposition (DED). The effects of neutron irradiation on microstructure (e.g., dislocations, loops, and nanoclusters), tensile properties (e.g., strength and ductility), and ion irradiation on Irradiation-Assisted Stress Corrosion Cracking (IASCC) will be explored. Results are collected from several projects supported by the Nuclear Science User Facilities (NSUF) program.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

Nitrogen effects on radiation response in 12Cr ferritic/martensitic alloys

Ferritic/martensitic steels are one of the best candidates for structural materials for high dose applications in next generation nuclear reactors. The composition of structural materials must be optimized for reliable service during irradiation. This study reports the effect of interstitial nitrogen on radiation response in 12Cr ferritic/martensitic HT9 steels having a controlled nitrogen concentration. Results show that a high amount of ‘free’ nitrogen in the matrix stabilizes the interstitial clusters which leads to (i) larger loop sizes (ii) lower loop density and (iii) slightly reduced radiation induced hardening. It also affects diffusion mechanism of Ni and formation of Ni/Si-rich precipitates.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In-situ observation of nano-oxide and defect evolution in 14YWT alloys

Nanostructured ferritic alloys (NFAs) are considered as candidates for structural components in advanced nuclear reactors due to their excellent radiation resistance as a result of a high density of nano-oxides (NOs) in the microstructure. Therefore, gaining an understanding on the stability of NOs under irradiation is crucial. In this study, we have investigated the evolution of defects and NOs in 14YWT NFAs under in-situ Kr ion irradiation at room temperature (RT) and 450°C up to 10 dpa. It has been found that irradiations at 450°C do not create any changes in the NOs, similar to the bulk irradiations. On the other hand, elemental mapping indicates that NOs dissolve mostly after 10 dpa irradiations at RT. Thus, while defects are both annihilated and pinned by NOs at low doses (before the dissolution of NOs), glissile loops start to escape to the foil surface at high doses (after the dissolution of NOs), justifying the significantly low fraction of (111) loops compared to the literature values. Furthermore, high resolution transmission electron microscopy analysis has shown that the NOs are mostly coherent Y 2 Ti 2 O 7 particles with pyrochlore crystal structure after both RT and 450°C irradiations, similar to those observed before irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Progression of creep deformation from grain boundaries to grain interior in Al-Cu-Mn-Zr alloys

Creep mechanisms are studied in θ'-Al 2 Cu-strengthened Al-Cu-Mn-Zr alloys at 300 and 350°C for (i) ACMZ, a base alloy without further alloying elements and (ii) RR350, a commercial alloy with additions of Ni and Co forming distinct grain-boundary precipitates. At high stresses, creep is dominated by dislocations bypassing θ' precipitates within grains via the Orowan mechanism, as evidenced by (i) very high stress exponent (n~20-25) and (ii) α-Al and θ' lattice strains (measured via in-situ neutron diffraction) evolving during creep in a manner consistent with load transfer from the plastically-deforming α-Al matrix to elastically-deforming θ' precipitates. At intermediate stresses, both alloys exhibit a n~3 regime, where α-Al and θ' lattice strains scale near-linearly with applied stress while remaining largely unaffected by strain accumulation, indicating that Orowan looping or dislocation pile-up around θ' is now inactive within the grains. Rather, dislocation motion occurs solely in θ'-precipitate-free zones (θ'-PFZ) where high dislocation densities are observed via TEM after creep deformation. Plastic flow at θ'-PFZ and/or localized pipe diffusion are expected to enable grain-boundary sliding (GBS), which is proposed as the rate-limiting mechanism in the n~3 regime. Ni/Co-rich precipitates at RR350 grain-boundaries, with negligible θ'-PFZ around them, share load (as determined via neutron diffraction) with the α-Al matrix more effectively than θ-Al 2 Cu precipitates at ACMZ grain-boundaries, with wide surrounding θ'-PFZ. So, high creep resistance in the n~3 GBS regime of RR350 is enabled by coarsening-resistant grain-boundary precipitates, forming without concomitant development of weak θ'-PFZ, which effectively share load with the grains.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A microscopic and crystallographic study of proton irradiated alloy 718

Solution annealing and age hardening are important processes for achieving good engineering and chemical properties of alloy 718. The composition of alloy 718 also plays an important role as it can affect both mechanical (γ”- and γ’-phase formation) and irradiation behaviors. Therefore, in this study five different sets of alloy 718 samples with two different chemical compositions were fabricated using different processing conditions and irradiated up to 4 dpa. Based on the effects of irradiation on the microstructural data of γ”- and γ’-phases, irradiation-induced dislocation statistics, and crystallographic data, it can be inferred that the samples processed at high solution annealing temperature (1093°C) behave better under irradiation compared to the samples processed at low solution annealing temperatures (945 and 1065°C).

36 MATERIALS SCIENCE↗
DOE OSTI2025

Grain boundary metastability controls irradiation resistance in nanocrystalline metals

Grain boundaries (GBs) in polycrystalline materials are powerful sinks for irradiation defects. While standard theories assume that a GB’s efficiency as a sink is defined solely by its character before irradiation, recent evidence conclusively shows that the irradiation sink efficiency is a highly dynamic property controlled by the intrinsic metastability of GBs under far-from-equilibrium irradiation conditions. In this paper, we reveal that the denuded (i.e., defect-free) zone, typically the signature of a strong sink, can collapse as irradiation damage accumulates. We propose a radiation damage evolution model that captures this behavior based on the emergence of a series of irradiation defect-enabled metastable GB microstate changes that dynamically alter the ability of the GB to absorb further damage. We show that these microstate changes control further defect absorption and give rise to the formation of a defect network that manifests itself as a net Nye-tensor signal detectable via lattice curvature experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Variations in γ′ formers and refractory elements for enhanced creep resistance and phase stability of an advanced Ni-based superalloy

The strong demands on increasing fuel efficiency have continuously driven the optimization of superalloys for high-performance applications. In this study, modifications to the chemistry and heat treatment of HAYNES® 282® alloy (H282) were performed by varying γ′ formers and refractory elements. It was found that increasing Ti and substituting W for some of the Mo in the newly designed alloy (Q) resulted in a significant improvement of creep resistance, up to 130% increase in creep life, compared to standard H282. It was found that Orowan loops and dislocation climb were the dominant creep deformation mechanisms in alloy Q, while extensive dislocation tangling as an additional configuration was observed in the baseline alloy. Moreover, phase stability investigations for up to 5000 h at 800 °C and 900 °C revealed a reduced formation of detrimental σ and μ phases in alloy Q when compared to H282. Atom-probe tomography (APT) revealed that the formation and growth of those phases were responsible for a decrease in Mo content in the matrix, thereby leading to a decrease in solid-solution strengthening in H282 over time. Furthermore, the coarsening of γ′ precipitates was retarded by the substitution of W for Mo, particularly under creep stress. The theoretical and experimental understanding of precipitation strengthening unraveled that higher optimal strengthening occurs at larger particle size for alloy Q, compared to the commercial formulation, further explaining the origin of enhanced creep resistance in the modified alloy.

36 MATERIALS SCIENCE↗