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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 1,729 records · Page 96
All sourcesNASA NTRSDOE OSTI
NASA NTRS2011

In Situ Strategy of the 2011 Mars Science Laboratory to Investigate the Habitability of Ancient Mars

The ten science investigations of the 2011 Mars Science Laboratory (MSL) Rover named "Curiosity" seek to provide a quantitative assessment of habitability through chemical and geological measurements from a highly capable robotic' platform. This mission seeks to understand if the conditions for life on ancient Mars are preserved in the near-surface geochemical record. These substantial payload resources enabled by MSL's new entry descent and landing (EDL) system have allowed the inclusion of instrument types nevv to the Mars surface including those that can accept delivered sample from rocks and soils and perform a wide range of chemical, isotopic, and mineralogical analyses. The Chemistry and Mineralogy (CheMin) experiment that is located in the interior of the rover is a powder x-ray Diffraction (XRD) and X-ray Fluorescence (XRF) instrument that provides elemental and mineralogical information. The Sample Analysis at Mars (SAM) suite of instruments complements this experiment by analyzing the volatile component of identically processed samples and by analyzing atmospheric composition. Other MSL payload tools such as the Mast Camera (Mastcam) and the Chemistry & Camera (ChemCam) instruments are utilized to identify targets for interrogation first by the arm tools and subsequent ingestion into SAM and CheMin using the Sample Acquisition, Processing, and Handling (SA/SPaH) subsystem. The arm tools include the Mars Hand Lens Imager (MAHLI) and the Chemistry and Alpha Particle X-ray Spectrometer (APXX). The Dynamic Albedo of Neutrons (DAN) instrument provides subsurface identification of hydrogen such as that contained in hydrated minerals

Mahaffy, Paul R.↗
DOE OSTI2018

Nano-scale microstructure damage by neutron irradiations in a novel Boron-11 enriched TiB 2 ultra-high temperature ceramic

Ultra-high temperature transition-metal ceramics are potential candidates for fusion reactor structural and plasma-facing components. We reveal the irradiation damage microstructural phenomena in Boron-11 enriched titanium diboride (TiB 2 ) using mixed-spectrum neutron irradiations, combined with state-of-art characterization using transmission electron microscopy (TEM) and high resolution TEM. Irradiations were performed using High Flux Isotope Reactor at ~220 and 620 °C up to 2.4 x 10 25 n.m -2 (E > 0.1 MeV). The calculated dose including contribution from residual Boron-10 ( 10 B) transmutation recoils, was ~4.2 displacements per atom. TiB 2 is susceptible to irradiation damage in terms of dislocation loop formation, cavities and anisotropic swelling induced micro-cracking. At both 220 and 620 °C, TEM revealed dislocation loops on basal and prism planes, with nearly two orders of magnitude higher number density of prism-plane loops. HRTEM, electron diffraction and relrod imaging revealed additional defects on {10$ \overline{1}\ $0} planes, identified as faulted prism-plane dislocation loops. High defect cluster density on prism planes will induce a-lattice parameter swelling of TiB 2 , as reported in literature. This anisotropic lattice parameter swelling induced grain boundary micro-cracking, the extent of which decreased with increasing irradiation temperature. The dominance of irradiation defect clusters on prism planes is different than typical hexagonal ceramics where dislocation loops predominantly form on basal planes, causing c-lattice parameter swelling. Helium generation and temperature rise from the transmutation of residual 10 B resulted in matrix and grain boundary cavities for the irradiation at 620 °C. In conclusion, the study additionally signifies isotopic enrichment as a viable approach to produce transition-metal diborides for potential nuclear structural applications.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Design and Development of Novel Equiatomic Refractory Multi Principal elemental Alloys Based on MoNbTi System for Use in Irradiation Environments

Multi Principal Elemental Alloys (MPEA) have emerged as promising materials for next-generation nuclear reactors due to their exceptional irradiation resistance. Eight equiatomic MPEA based on the MoNbTi system, comprising of elements with low thermal neutron absorption cross-sections were explored using a combined approach employing empirical parameter estimations, and CALPHAD simulations by which the phases and elemental segregation observed in all the alloys in the as-cast state were predicted. Solution heat treatment at 1500°C transformed five alloys into single-phase matrix materials, enhancing homogeneity and reducing hardness. The densities of the alloys ranged between 6.47 to 7.68 g/cm3, hardness between 472 and 656 VHN, Young’s modulus between 142 GPa to 169 GPa, shear modulus between 54 GPa and 62 GPa, bulk modulus between 117 GPa to 194 GPa and Poisson’s ratio between 0.3 to 0.35. The in-situ high temperature Xray diffraction results, differential scanning calorimetry and dilatometry results up to 1000°C suggested the high temperature phase stability of the MPEA. Subsequent ageing heat treatment at 800 and 1000 oC for 96 hours revealed significant secondary phase precipitation in MoNbTiZr, MoNbTiZrV, and MoNbTiCrA. Oxidation studies at 800 oC for 24 hours in air revealed superior oxidation resistance and cubic rate law dependence in Cr containing MPEA especially in MoNbTiCrAl, while severe mass gain resulting in total disintegration and exfoliation in MoNbTiZr and MoNbTiZrV. This comprehensive study underscores the potential of novel MPEA as promising materials for advanced nuclear reactor applications, shedding light on their microstructural control, mechanical properties, thermal stability, and oxidation resistance.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2008

MASSIM, the Milli-Arc-Second Structure Imager

The MASSIM (Milli-Arc-Second Structure Imager) mission will use a set of achromatic diffractive-refractive Fresnel lenses to achieve imaging in the X-ray band with unprecedented angular resolution. It has been proposed for study within the context of NASA's "Astrophysics Strategic Mission Concept Studies" program. Lenses on an optics spacecraft will focus 5-11 keV X-rays onto detectors on a second spacecraft flying in formation 1000 km away. It will have a point-source sensitivity comparable with that of the current generation of major X-ray observatories (Chandra, XMM-Newton) but an angular resolution some three orders of magnitude better. MASSIM is optimized for the study of jets and other phenomena that occur in the immediate vicinity of black holes and neutron stars. It can also be used for studying other phenomena on the milli-arc-second scale, such as those involving proto-stars, the surfaces and surroundings of nearby active stars and interacting winds.

Skinner, Gerry↗
DOE OSTI2025

Structure evolution and tin redistribution during oxidation of Zircaloy-4 at 500°C

Zirconium (Zr) alloys are widely used as fuel cladding in nuclear power reactors due to their thermal stability, mechanical durability, corrosion resistance, and low neutron absorption cross-section. However, their performance is challenged by oxidation in reactor environments, making the study of Zr alloy corrosion behavior crucial for ensuring the safety, longevity, and economic viability of nuclear power systems. While the oxidation behavior of Zr-based cladding materials has been extensively studied since the 1950s, a mechanistic understanding into the relationship between structure evolution, solute element redistribution, and properties remains elusive. Valuable insights may be obtained through advanced experimental methods, such as in-situ and high resolution microscopy techniques. Here, in this study, the oxidation behavior of Zircaloy-4 at 500 °C in O 2 is characterized using a multimodal advanced characterization approach. Using in-situ X-ray diffraction, the phase evolution from metastable to stable oxides is tracked in real time. Complementary high-resolution techniques, including electron microscopy and atom probe tomography, reveal nanoscale insights into the microstructural changes and solute redistribution across the oxide/metal interface. Nanohardness mapping across the oxide/metal interface highlights localized mechanical property variations that may be linked to changes in microstructure and crystal structure within the oxide layer. These findings offer valuable insights into the microstructure and property evolution of Zircaloy-4 during oxidation, contributing to a better understanding of microstructural changes in Zr-based alloys under oxidative environments.

APT↗
DOE OSTI2019

Effect of neutron and ion irradiation on the metal matrix and oxide corrosion layer on Zr-1.0Nb cladding alloys

A detailed study has been carried out on recrystallised Zr-1.0Nb alloys corroded and irradiated under different conditions, including ex-autoclave and ex-reactor samples. After 540 days in reactor and damage around 5 dpa, the neutron irradiated sample shows no serious evidence for radiation enhanced corrosion and is still in pre-transition stage. The good corrosion resistance of the neutron irradiated Zr-1.0Nb can be related to the higher volume fraction of tetragonal phase and fewer interconnected nano porosity/cracks in the oxide. This indicates less tetragonal to monoclinic transition, leads to more protective oxide in the neutron irradiated sample, containing little evidence for short circuit paths for the penetration of oxygen or water towards the metal-oxide interface. These observations on a sample with a slow overall oxidation rate are consistent with the hypothesis that interconnected porosity can lead to early transitions and rapid oxidation. Tetragonal oxide can be either stabilised by irradiation, or stabilised by local release of impurity species from SPPs such as dissolution of Fe from Zr-Nb-Fe precipitates or radiation introduced precipitates (RIPs) which is likely to be small β-Nb clusters. The oxide consists of well-aligned columnar-equiaxed microstructure in the autoclave sample while a more complex oxide grain structure was observed in the neutron-irradiated sample. As oxide continues to grow, there are more < a > and < c > loops, dissolved Fe and RIPs in the metal matrix, however, the corrosion rate is low enough for the tetragonal oxide to stabilise and suboxide + Zr(Osat) phases exist for protectiveness, so there is no enhanced corrosion after radiation. In situ ion irradiation in the TEM revealed no visible defect clusters or voids in the oxide, suggesting that radiation damage to the metal matrix rather than oxide may have a stronger effect on corrosion mechanisms after neutron irradiation, however, cascade damages are not visible in this case. Neutron irradiation also seems to have little effect on promoting fast oxidation or dissolution of β-Nb precipitates into the surrounding oxide or metal during irradiation. Furthermore, these results are discussed in the light of the current mechanisms for corrosion of nuclear fuel cladding alloys.

(S)TEM↗
DOE OSTI2020

Effects of Neutron Irradiation on the Micro/Nano Scale Structure and Fission Product Distribution of TRISO Coated Particle Fuel Kernels from AGR Experiments

The Advanced Reactor Technologies (ART) Program at Idaho National Laboratory (INL) includes the Advanced Gas Reactor (AGR) fuel development and qualification program that consists of fuel fabrication, experiment irradiations, post-irradiation examination (PIE) and safety testing to assess tristructural isotropic (TRISO) fuel performance during normal irradiation and under potential accident conditions. Advanced microscopy work on selected AGR-1 and AGR-2 unirradiated and irradiated fuel specimens is performed as part of the PIE effort. PIE work on fuel from the first experiment irradiation, AGR-1, began at INL in April 2010, and AGR-2 PIE began at INL in July 2014. This work scope includes University of Florida (UF) performing advanced electron microscopy examination and analysis using facilities at the Center for Advanced Energy Studies (CAES), INL or UF’s electron microscopy facilities. Electron microscopic examination and analysis may include scanning transmission electron microscopy (STEM), transmission electron microscopy (TEM), selected area diffraction (SAD), electron energy loss spectroscopy (EELS), electron dispersive spectroscopy (EDS) and atom probe tomography (APT) on the fuel kernels of TRISO-coated fuel particles. The electron microscopy lamellae will be provided by INL and will be available at CAES or Irradiated Materials Characterization Laboratory (IMCL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal conductivity evaluation of ion irradiated Si 3 N 4 and ZrN ceramics using spatial domain thermoreflectance

Nitride ceramics have been investigated for different applications in the nuclear industry, such as space nuclear power, fusion reactor diagnostics and plasma heating, inert matrix fuels, and accident tolerant fuels. Although thermal conductivity remains one of the most important properties to track following irradiation, traditional techniques such as laser flash and xenon flash are limited to bulk sample characterization, which requires lengthy and cost-consuming neutron irradiation. This work used spatial domain thermoreflectance (SDTR) for the micrometer-scale measurement of thermal conductivity in 15 MeV Ni ion-irradiated silicon nitride and zirconium nitride from 1 to 50 dpa and 300 to 700 °C. Here, the SDTR-measured unirradiated thermal conductivity was found to be consistent with the published data on bulk samples. Electrically conductive ZrN exhibits modest reduction after irradiation which is minimal at the highest irradiation temperatures. In electrically insulating Si 3 N 4 , the reduction is more significant and unlike ZrN, the reduction remains significant even at a higher irradiation temperature. The thermal resistance evolution following irradiation was compared with lattice swelling, which was determined using grazing incidence x-ray diffraction, and radiation-induced defects were observed using transmission electron microscopy. A saturation value was observed between 15 and 50 dpa for thermal conductivity degradation in both nitride ceramics and a direct correlation with high-temperature defect recombination was observed, as well as the potential presence of additional carrier scattering mechanisms.

36 MATERIALS SCIENCE↗
DOE OSTI2019

NEUTRON-IRRADIATION–INDUCED DAMAGE ASSESSMENT IN CONCRETE USING COMBINED PHASE CHARACTERIZATION AND NONLINEAR FAST FOURIER TRANSFORM SIMULATION

Understanding irradiation effects in concrete cavities of light water reactors is critical to the long-term operation (60+ years) of US nuclear power plants. Susceptibility of concrete to neutron irradiation greatly varies as a function of its constituents: coarse aggregates, sand, and hardened cement paste (hcp). In particular, higher irradiation–susceptibility of aggregates was found to be a direct function of radiation-induced volumetric expansion (RIVE), or the propensity of aggregates to swell as a function of mineral contents, structures and textures. Irradiation-induced amorphization is accompanied by significant swelling, especially in silicates. For example, the maximum volumetric expansions of quartz and feldspars have been shown to be as large as ≈ 18% and ≈ 8%, respectively, while the change of density in calcite remains rather low (≈ 0.3–0.5%). High swelling is associated with a high number of covalent bonds (Si–O) and a high degree of polymerization of [SiO4] tetrahedrons. Differential RIVEs between rock-forming minerals cause high stresses and may potentially develop cracking within the aggregates at the interface between aggregates and the hcp and within the hcp. Modeling neutron-irradiation effects in concrete requires addressing the problem at a scale that captures both the details of the aggregates’ mineralogy and texture, which is <~ 100 μm, and the representative size of concrete, which is ~ 10 cm. This requires high-resolution, large mapping capabilities. To this end, the US Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program developed an original tool suite called Microstructure Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) in which Nonlinear 2D-fast Fourier transform (FFT) techniques are employed to model mechanical or irradiation-induced cracking of high-resolution complex microstructures derived from combined mapping techniques, including micro x-ray fluorescence (m-XRF), energy-dispersive x-ray spectroscopy (EDS), electron backscatter diffraction (EBSD), ellipsometry, and scanning electron microscopy (SEM). Irradiated mineral properties are derived from the Irradiated Minerals, Aggregate and Concrete (IMAC) database, which includes available relevant literature data. Rigorous coupled damage-creep simulations are obtained through sequential linear analysis. Illustrations of actual accelerated irradiation-damage propagation in a complex microstructures are presented herein.

Le Pape, Yann↗
DOE OSTI2020

Characterising Ion-Irradiated FeCr: Hardness, Thermal Diffusivity and Lattice Strain

Ion-irradiated FeCr alloys are useful for understanding and predicting neutron damage in the structural steels of future nuclear reactors. Previous studies have largely focused on the structure of irradiation induced defects, probed by transmission electron microscopy (TEM), as well as changes in mechanical properties. Across these studies, a wide range of irradiation conditions has been employed on samples with different processing histories, which complicates the analysis of the relationship between defect structures and material properties. Furthermore, key properties, such as irradiation-induced changes in thermal transport and lattice strain, are little explored. Here we present a systematic study of Fe3Cr, Fe5Cr and Fe10Cr binary alloys implanted with 20 MeV Fe3+ ions to nominal doses of 0.01 dpa and 0.1 dpa at room temperature. Nanoindentation, transient grating spectroscopy (TGS) and X-ray micro-beam Laue diffraction were used to study the changes in hardness, thermal diffusivity and strain in the material as a function of damage and Cr content. Our results suggest that Cr leads to an increased retention of irradiation-induced defects, causing substantial changes in hardness and lattice strain. However, thermal diffusivity varies little with increasing damage and instead degrades significantly with increasing Cr content in the material. We find significant lattice strains even in samples exposed to a nominal displacement damage of 0.01 dpa. The defect density predicted from the lattice strain measurements is significantly higher than that observed in previous TEM studies, suggesting that TEM may not fully capture the irradiation-induced defect population. (C) 2020 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

FeCr alloys↗
DOE OSTI2025

Structural characterization of highly alloyed (Al,Gd)N thin films

Highly alloyed (Al,Gd)N is of potential interest in a variety of applications, including neutron detection and in devices such as non-volatile memory. Gd has been shown to have very low equilibrium solubility in AlN at room temperature; however, non-equilibrium deposition techniques such as sputtering are able to deposit thin films, which incorporate large amounts of Gd. Here, we characterize a highly-alloyed (Al,Gd)N combinatorial thin film grown by RF sputtering on a GaN substrate, looking for any evidence of chemical or phase segregation or structural disorder in the films. Compositions with between 13% and 32% Gd (on a cation basis) were studied. No evidence was found for chemical or phase segregation in any studied composition. Higher degrees of Gd incorporation led to greater structural disorder in the film and a tendency toward amorphization; however, electron diffraction shows that the film does not become fully amorphous at any of the studied compositions, instead retaining textured local order even at 32% Gd. Electron energy loss spectra suggest that the material retains a locally wurtzite-like tetrahedral bonding environment at all studied compositions.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperature of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and 500 oC, respectively. The coating layer contains both carbon and vanadium elements as quantified by WED, and the phases mainly consist of a mixture of V2C and VC, which was confirmed using X-ray diffraction patterns. In addition, the ratio between V and C varies with processing temperature, and it was observed that a higher temperature promotes the carbon adsorption and increases thickness of the coating. With optimized deposition conditions, we can apply the coating technique toward the actual T91 cladding materials, and provide the possibilities for the real application in sodium-cooled fast reactors (SFRs). Diffusion couple experiments were performed at both 550 oC and 690 oC, which corresponds to normal and aggressive operating temperatures, respectively. The results show that vanadium carbide coating with wider thickness (8 µm) and lower carbon concentration (27 at.%) reduced the width of the inter diffusion region, indicating that vanadium carbide coating can mitigate FCCI effectively. In specific, inter-diffusion between Fe and Ce was prohibited over most area, but Ce diffusion occurred toward the coating and the Fe substrate through thinner coating layer, which needs further optimization in terms of uniform coating thickness. Overall, it is concluded that this coating process can be successfully applied onto the inner surface of HT9 cladding tubes and the FCCI can be effectively mitigated if not totally eliminated.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Impact of Backbone Rigidity on the Thermomechanical Properties of Semiconducting Polymers with Conjugation Break Spacers

There remains a lack of fundamental understanding in theroleofbackbonerigidityonthe thermomechanical properties of conjugated polymers. Here, we provide the first holistic approach to understand the fundamental influence of backbone rigidity on an n-type naphthalene diimide-based conjugated polymer, denoted by PNDI-Cx, through insertion of a flexible conjugation break spacer (CBS). CBS lengths are varied from fully conjugated with zero alkyl spacer (PNDI-C0) to a seven- carbon alkyl spacer (PNDI-C 7 ), with the CBS engineered into each repeat unit for systematic evaluation. Solution small-angle neutron scattering and oscillatory shear rheometry were employed to provide the first quantitative evidence of CBS influence over conjugated polymer chain rigidity and entanglement molecular weight (M e ), demonstrating a reduction in the Kuhn length from 521 to 36 Å for fully conjugated PNDI-C0 and PNDI-C6, respectively, as well as a nearly consistent Me of ~15 kDa upon the addition of CBS. Thermomechanical properties, such as elastic modulus and glass-transition temperature, were shown to decrease with an increasing length of CBS. An extraordinary ductility, upwards of 400% tensile strain before fracture, was observed for high-molecular-weight PNDI-C4, which we attribute to a high number of entanglements and disruption of crystallization. Furthermore, the deformation mechanism for PNDI-Cx was studied under strain through X-ray diffraction, polarized UV-vis spectroscopy, and atomic force microscopy. Overall, this work sheds light on the important role of backbone rigidity in designing flexible and stretchable conjugated polymers.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

An anomalous structure disordering in Zr–Cu–Ag supercooled glass-forming liquids

An anomalous ordering-to-disordering transition in a Zr–Cu–Ag supercooled glass-forming liquid has been revealed by in situ high-energy x-ray diffraction. Upon heating, the supercooled liquid exhibits a structural crossover at T c = 716 K, evident by the change in statistical characteristics of the first sharp diffraction peak. When annealed in the vicinity of T c , the supercooled liquid first relaxes from its initial state to a denser and more ordered state. Subsequently, it becomes disordered at the medium-range scale with a slowing down of densification, as demonstrated by the broadening of the first sharp diffraction peak and a decrease in the peak intensities of the higher atomic shells in the reduced pair distribution function. Finally, a scenario for the evolution of less correlated locally favored clusters in the liquid–liquid phase transition is discussed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Electron Microscopic Examination of an Irradiated TRISO Coated Particle of AGR-2 Experiment: AGR2-222-RS019

The transmission electron microscopic (TEM) examination of an irradiated tristructural isotropic (TRISO) coated particle from advanced gas reactor (AGR) 2 fuel Compact 2-2-2 is presented in this report. Compact 2-2-2 refers to the compact in Capsule 2 at Level 2 of Stack 2. The examination on AGR2 222 RS019 is focused on the interface between the silicon carbide (SiC) to inner pyrolytic carbon (IPyC) layer, central part of SiC, and the interface between SiC and outer pyrolytic carbon layer. The SiC grain boundary distribution and associated fission product precipitates were characterized by scanning transmission electron microscopy (STEM) energy dispersive spectroscopy (EDS), and precession electron diffraction (PED) in the transmission electron microscope. Two separate locations on SiC layer were selected for on the basis of the intact of the buffer layer in the irradiated TRISO particle. The key observation in this particle is that the compositional nature of grain boundary precipitates is more complex near IPyC/SiC layer where Pd, Pd-Ag, Pd-U containing precipitates can be observed. However, the compositions of precipitate towards outer most region of SiC mainly contain Pd. Ag has been observed in the IPyC/SiC layer in one of the two locations. Comparison of the results to previously irradiated and analyzed, Variant 3, safety tested AGR-1 TRISO particles (AGR1-433-001 and AGR1 222-RS004) showed the grain boundary distributions have some noticeable differences. The differences in coincident site lattice grain boundary and misorientation angles are possibly inherent to the fabrication process. Neutron irradiation induced void distribution has been found to be nonuniform. A high concentration of smaller voids is found at stacking faults compared to those in the SiC matrix. Larger void sizes are observed in the SiC location that is near to the broken buffer layer. This void size variation with integrity of buffer layer can potentially affect the fission product retention. However, the average void size continuously decreases along the radius of SiC in both locations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Fabrication of yttrium hydride for high-temperature moderator application

We report given the superior thermal stability and highly attainable hydrogen density, yttrium hydride is an excellent high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. Large-scale, crack-free, bulk yttrium hydride is in high demand; however, fabrication of yttrium hydride is challenging and has not been demonstrated for nearly half century. The associated challenges are discussed herein. In response to these challenges, a hydriding system was designed and constructed at Oak Ridge National Laboratory and was used to successfully fabricate crack-free yttrium hydride in complex geometries at large scales. This was accomplished by precisely controlling the hydrogen’s partial pressure and the retort temperature, which was informed by the well-established thermodynamic properties of the binary H–Y system. Hydrogen content in as-fabricated hydride was determined by weight change, an approach which was considered reliable based on the use of ultra–high-purity yttrium, the absence of oxide phases up to levels detectable using x-ray diffraction (XRD), and the significant weight gain. Hydrogen distribution along one yttrium hydride rod was evaluated with XRD analysis on materials extracted from different locations on the rod. The results indicated a relatively homogeneous hydrogen distribution along the hydride rod, with <3% uncertainty in the fraction of the δ-phase hydride. In addition, significant efforts are being dedicated to establish a complete database summarizing the thermomechanical and physical properties of as-fabricated yttrium hydride and the irradiation response to facilitate its deployment as a high-temperature moderator in advanced nuclear reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Complete the Design of Benchtop Fiber-Optic Based Photo-Thermal Radiometry System

Thermal transport properties of nuclear fuels, i.e., thermal conductivity and diffusivity, are critical parameters in the thermal design of a nuclear reactor. It is well known that thermal transport properties change temporally and spatially due to dynamic microstructural evolution as a result of neutron irradiation. A real-time instrument to monitor the local thermal property change, which can survive the harsh in-pile environment, will give researchers unprecedented access to fuel behavior during irradiation. Laser-based photothermal techniques are an ideal candidate to accomplish this specific task, because of the remote, non-contact, and non-destructive features. In this report, we present the progress in the development of a bench-top photothermal radiometry (PTR) system to measure the thermal diffusivity of solid samples, and the preliminary design of a fiber-based PTR instrument to measure the real-time, in-pile thermal diffusivity of nuclear fuels. The PTR technique measures the local temperature response from a modulated heat source by detecting the sample radiation heat flux. Initially a well-developed thermal wave analytical model was fit to the data to obtain the thermal diffusivity. However, systematic errors from the diffraction effect and non-linearity issues resulted in thermal diffusivity measurements higher than literature values. These effects have been studied and an improved analytical model has been developed, and the physics behind the phenomena are still under investigation. A procedure for optimizing measurements from the bench-top system has been provided. The experiences from the bench top measurement were then used to guide the fiber-based system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Synthesis and Thermal Oxidation Resistance of Boron-Rich Boron–Carbide Material

A boron-rich boron–carbide material (B4+δC) was synthesized by spark plasma sintering of a ball-milled mixture of high-purity boron powder and graphitic carbon at a pressure of 7 MPa and a temperature of 1930 °C. This high-pressure, high-temperature synthesized material was recovered and characterized by X-ray diffraction, X-ray photoelectron spectroscopy, Raman spectroscopy, Vickers hardness measurements, and thermal oxidation studies. The X-ray diffraction studies revealed a single-phase rhombohedral structure (space group R-3m) with lattice parameters in hexagonal representation as a = 5.609 ± 0.007 Å and c = 12.082 ± 0.02 Å. The experimental lattice parameters result in a value of δ = 0.55, or the composition of the synthesized compound as B4.55C. The high-resolution scans of boron binding energy reveal the existence of a B-C bond at 188.5 eV. Raman spectroscopy reveals the existence of a 386 cm−1 vibrational mode representative of C-B-B linear chain formation due to excess boron in the lattice. The measured Vickers microhardness at a load of 200 gf shows a high hardness value of 33.8 ± 2.3 GPa. Thermal gravimetric studies on B4.55C were conducted at a temperature of 1300 °C in a compressed dry air environment, and its behavior is compared to other high-temperature ceramic materials such as high-entropy transition metal boride. The high neutron absorption cross section, high melting point, high mechanical strength, and thermal oxidation resistance make this material ideal for applications in extreme environments.

36 MATERIALS SCIENCE↗