SEARCH · Engineering Papers

Results for “uranium (U)”

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.

SourceResearch connectionFrom yearThrough year

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

At least 433 records · Page 24

All sources NASA NTRS DOE OSTI

DOE OSTI2020

Synthesis and Structure of Uranium‐Silylene Complexes

Abstract While carbene complexes of uranium have been known for over a decade, there are no reported examples of complexes between an actinide and a „heavy carbene.“ Herein, we report the syntheses and structures of the first uranium‐heavy tetrylene complexes: (CpSiMe 3 ) 3 U‐Si[PhC(NR) 2 ]R’ (R= t Bu, R’=NMe 2 1 ; R= i Pr, R’=PhC(N i Pr) 2 2 ). Complex 1 features a kinetically robust uranium‐silicon bonding interaction, while the uranium‐silicon bond in 2 is easily disrupted thermally or by competing ligands in solution. Calculations reveal polarized σ bonds, but depending on the substituents at silicon a substantial π‐bonding interaction is also present. The complexes possess relatively high bond orders which suggests primarily covalent bonding between uranium and silicon. These results comprise a new frontier in actinide‐heavy main‐group bonding.

Brackbill, I. Joseph↗

DOE OSTI2021

Thermophysical Properties of Mixtures of Thorium and Uranium Nitride

Abstract The miscibility, lattice parameter, and thermophysical properties of (Th 0.2 U 0.8 )N and (Th 0.5 U 0.5 )N have been investigated. It is shown that additions of thorium nitride (ThN) to uranium nitride (UN) increases the thermophysical performance of the mixed nitride fuel form in comparison to reference UN. In the more dilute limit, additions of ThN serve as a burnable neutronic poison and reduces the change in k eff over the lifecycle of the fuel. At higher concentrations, additions of ThN serve as a significant fertile source of 233 U. Where appropriate, comparisons to previous work on UN + PuN mixtures are made, as this is a comparable fuel form for potential fast reactor concepts, and a suitable point of contrast in the possible design space afforded by mixed (Th x U 1 − x )N fuel forms. The data from this work are the input parameters for finite element modeling of the temperature distribution in a compact reactor. The results of modeling and simulation of this core design are shown for the case of steady-state operation and during double, adjacent heat pipe failure.

Parker, S. S. (ORCID:000000018597440X)↗

DOE OSTI2024

Credible Criticality Safety Margin in the 30B Package with LEU+ UF 6 and Hypothetical Water Ingress

The commercial nuclear industry is pursuing advancements in fuel and reactor design that increase the uranium enrichment above 5 wt. % 235 U. These advancements will necessitate the ability to transport bulk quantities of UF 6 at increased enrichments. Currently, the 30B cylinder is the primary container used by the industry for UF 6 storage and transportation and has a long history of successful shipments. This container can support up to 2,277 kg of UF 6 at a maximum enrichment of 5 wt. % 235 U. Previous evaluations have assessed the potential impact of criticality safety for 30B transport at higher enrichments but assumed moderator intrusion would not require evaluation. Although current regulations allow for the exception of moderator intrusion for UF 6 packages through the design and quality control of the package content, this exception is limited to enrichments up to 5 wt. % 235 U. Thus, an investigation of moderator intrusion into a 30B cylinder should be performed. Moderator intrusion into a 30B cylinder is a unique condition for criticality safety evaluation in transportation because of the violent chemical reactions that occur between UF 6 and H 2 O. The resulting intrusion is strongly dependent on the breach size, breach location, breach interface solid/ullage of the UF 6 content, the UF 6 distribution (which is temperature dependent), and temperature/pressure conditions which are dynamic in accident conditions. Additionally, the complexity of the HF-UO 2 F 2 -H 2 O interface that occurs during the event can influence the potential solubility of uranium in the system, as well as influence the amount of UF 6 reaction with H 2 O. With the operating experience from Orano Federal Services LLC, the UF 6 chemical expertise from Oak Ridge National Laboratory (ORNL), and the criticality safety expertise from ORNL, this paper evaluates the neutronic conditions (i.e., k eff ) that apply the understood chemistry and experimental conditions that occur during moderator intrusion of a breached 30B cylinder under postulated accidents. This report examines the historical evaluations of UF 6 transport and expands these evaluations for the enrichments expected for nuclear industry advancement. These simulations primarily feature a homogeneous mixture of UF 6 and H 2 O as an infinite media system and as a sphere with water reflection given an impurity limit of 0.5 wt. % UF 6 . This report demonstrates that this H/U limit is valid up to 8 wt. % enrichment for both H 2 O and HF as the moderating mixture. The water-reflected homogeneous mixed spheres evaluated demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. Additionally, this report evaluates moderator ingress scenarios of a 30B cylinder. The 30B cylinder simulations vary the cylinder orientation, the mass of UF 6 in the system, the mass of H 2 O in the system, and how much H 2 O has reacted with UF 6 . The moderator ingress of a 30B cylinder was evaluated for the water-reflected homogeneous mixed spheres to demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. These simulations are standard practice and are independent of the UF 6 -H 2 O reaction. This work also explored more complex simulations that layer the H 2 O over the UF 6 , which incorporates the amount of reacted UF 6 as a separate layer and considers solubility limits of UO 2 F 2 in H 2 O as it is produced. This approach is intended to simulate the postulated event of a large hairline crack occurring on a submerged cylinder. Finally, this work considers the potential of H 2 O mixing into the UF 6 solid heterogeneously using a sponge-like model. The water ingress into the UF 6 is treated as random size spheres, and the reaction products form into a layer over the spheres.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Probing the hydrolytic degradation of UF 4 in humid air

This manuscript describes the chemical transformations that occur during hydrolysis of uranium tetrafluoride (UF 4 ) due to its storage in humid air (85% and 50% relative humidity) at ambient temperatures. This hydrolysis was previously reported to proceed slowly or not at all (depending on the percent relative humidity); however, previous reports relied primarily on X-ray diffraction methods to probe uranium speciation. Here, we employ a battery of physiochemical probing techniques to explore potential hydrolysis, including Raman spectroscopy, powder X-ray diffraction, 19 F nuclear magnetic resonance spectroscopy, scanning electron microscopy, and focused ion beam microscopy with energy-dispersive X-ray spectroscopy. Of these, only Raman spectroscopy proved to be particularly useful at observing chemical changes to UF 4 . It was found that anhydrous UF 4 slightly oxidizes over the course of thirteen days to Schoepite-like uranium complexes and possibly UO 3 . In contrast, UF 4 exposed to 50% relative humidity slightly decomposes into UO 2 F 2 , Schoepite-like uranium complexes, and possibly a high order uranium oxide that eluded chemical assignment (U x O y ). Despite the rich chemical speciation observed in our Raman spectroscopy measurements, X-ray diffraction and 19 F NMR measurements on the same material showed no changes. Microscopy measurements suggest that the observed reactions between UF 4 and water occur primarily on the surface of UF 4 particulates via a method that is visually similar to surface corrosion of metals. Therefore, we postulate that NMR spectroscopy and X-ray diffraction, which are well-suited for bulk analysis, are less suited than Raman spectroscopy to observe the surface-based reactions that occur to UF 4 when exposed to humid air. Considering the importance of UF 4 in the production of nuclear fuel and weapons, the results presented herein are widely applicable to numerous nuclear science fields where uranium detection and speciation in humid environments is of value, including nuclear nonproliferation and nuclear forensics.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Improving stability of ALD ZrN thin film coatings over U-Mo dispersion fuel

Atomic layer deposition (ALD) of ZrN is a candidate technology for coating U-Mo dispersion fuel as a diffusion barrier coating. During the early development stages of the coating the ZrN deposition showed to be mechanically unstable and ultimately resulted in spalling. Based on experimental outcomes it was found that instabillity can be eliminated through the introduction of a thin amorphous (a) Al 2 O 3 interlayer coating deposited in between the ~1-μm thick ZrN and the U-Mo. substrate. To elucidate these findings simulations were performed with density functional theory (DFT) to measure work of adhesions at different interfaces while finite element modeling (FEM) was performed to measure the residual stress distribution. DFT indicated that the ZrN coating deposited over (a)-Al 2 O 3 is ~2.5 times stronger when compared to direct deposition over U or UO 2 substrate. While calculations from FEM recognized; (1) large stress concentrations can originate from the irregularly distributed native surface oxides (UO 2 ) over U-Mo and (2) stress concentrations can be reduced if those surface UO 2 can be modified to a uniform layer (achieved after application of a 8 nm (a)-Al 2 O 3 interlayer). These results provided explanations and confirmed the role played by interlayer in eliminating the original mechanical instability of the ZrN.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Mapping the Microhardness of Al Matrix in U-7Mo/Al Dispersion Fuels at Medium and High Burn-up by in situ Nanoindentation

The mechanical degradation of the Al matrix in U-7Mo dispersion fuels was detected after irradiation in previous studies, but more details of the degradation mechanism are not clear yet to clarify the root cause. In this study, microhardness mapping in fresh and irradiated U-7Mo dispersion fuels coated with ZrN, with burn-up of 3.35 × 10 21 and 6.28 × 10 21 f/cm 3 , are measured to reveal the hardening of the Al matrix as a function of distance to fuel particles. Micro cubes were pulled by plasma focused ion beam scanning electron microscopy and tested by in situ nanoindenter with Berkovitch tip. Size effects are measured in medium and high burn-up samples. The higher burn-up specimen exhibits a more pronounced size effect when indent load is lower than 20 mN. Size effect correction models are also calculated by plots fitting, which predicts the true microhardness without size effect. Results shows that the hardness of the Al matrix near the ZrN coating has the highest value, then decreases moving away from the coating and become stable when the distance reaches ∼10 µm. The hardness starts increasing again getting closer to the next fuel particle, which makes the hardness distribution between two fuel particles to be “U” shape.

36 - MATERIALS SCIENCE↗

DOE OSTI2026

Transient Triamidoamine Neptunium(V)–Mono(Imido) Complexes: C–H Activations and Hydrogen Atom Transfer Driven by Effective Nuclear Charge

Metal-mono(imido) linkages have been known for seven decades, and they are found in transition metal, main group, lanthanide, thorium, and uranium complexes. However, transuranium-mono(imido) complexes remain unknown in any scenario. Here, we present evidence for transient neptunium(V)–mono(imido) complexes. Treatment of [Np III (Tren TIPS )] (1, Tren TIPS = {N(CH 2 CH 2 NSiPr i 3 ) 3 } 3– ) with N 3 R (R = SiMe 3 ; 1-adamantyl, Ad) results in N 2 evolution and dark purple solutions consistent with the formation of [Np V (Tren TIPS )(NR)] (3NpNR). However, solutions of 3NpNR rapidly turn orange, where for R = SiMe 3 the isolated 1:1 products are [Np IV (Tren TIPS ){N(H)SiMe 3 }] (4a) and [Np IV (Tren TIPS-2H ){N(H)SiMe 3 }] (4b, Tren TIPS-2H = {N(CH 2 CH 2 NS i Pri 3 ) 2 (NCH 2 CH 2 NSiPr i 2 C[Me]=CH 2 )} 3– ). The latter contains a dehydrogenated-Pr i vinyl functionality accounting for the source of the two amido H atoms. The reaction for R = Ad proceeds similarly, but only [Np IV (Tren TIPS ){N(H)Ad}] (5a) could be unequivocally confirmed, though its isolation suggests generality of the imido-to-amido functional group transformation. Complexes 4a/4b exhibit slow relaxation of their magnetization, adding to the small number of transuranium single ion magnets. Experimental and computational analysis suggests that the amido products are formed by C–H activation and two sequential hydrogen atom transfer reactions involving a three-step proton-coupled electron-transfer sequence of H • radical abstraction, electron transfer, then another H • radical abstraction step. In contrast to transient 3NpNR, the 5f 2 uranium(IV)-imido complex [K(2.2.2-cryptand)][U IV (Tren TIPS )(NSiMe 3 )] (8UNSiMe 3 ) is robust, even in boiling THF, suggesting the transience of 5f 2 3NpNR is not due to the 5f n -count but the increased effective nuclear charge of neptunium vs uranium. This work highlights divergence of uranium- and neptunium-imido stabilities, emphasizing that the latter is an inherently challenging synthetic target.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Characterization and Optimization of a Spectral Window for Direct Gaseous Uranium Hexafluoride Enrichment Assay Using Laser-Induced Breakdown Spectroscopy

Through a systematic scanning of 235 U and 238 U emission lines between 280 nm and 745 nm, the optimal emission line for direct gaseous uranium hexafluoride (UF 6 ) enrichment assay using laser-induced breakdown spectroscopy (LIBS) was found. Screening for spectral features that are potentially useful for U isotopic analysis was gauged from the magnitude of the 235 U– 238 U isotopic shift and the signal-to-background ratio of the emission line through a parameter termed ΔSBR 235U–238U . The ΔSBR spectrum shows peaks at wavelength positions where there are strong lines with significant 235 U– 238 U shifts. The screening identified 13 spectral-window candidates, which were down selected based on their overall accuracy in predicting the 235 U enrichment of three UF 6 samples of natural (0.720 atom% 235 U) and low-enriched (4.675 atom% and 9.157 atom% 235 U) grades. The U(I) 646.498 nm emission line, with a determined 235 U– 238 U isotopic shift of -17.7 pm, was found to be the optimal spectral window for direct UF 6 enrichment assay. The root mean square error for enrichment assays on the three natural and low-enriched UF 6 samples, with each sample measured in six replicates, was 0.31% in absolute 235 U content. Each measurement comprised LIBS signals accumulated from 3000 laser shots. In conclusion, the analytical bias and precision were better than 0.5% and 0.3%, respectively, in absolute [ 235 U/( 235 U + 238 U)] ratios. Specific for the two low-enriched UF 6 samples, the relative standard deviations from six replicated measurements were around 2%.

47 OTHER INSTRUMENTATION↗

DOE OSTI2023

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235 U content. Additionally, finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm 2 . No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Direct Extraction of Uranium Oxides with N,N -Di(2-ethylhexyl)isobutyramide

The direct extraction of uranium from voloxidized nuclear fuel into an organic solvent offers several potential advantages over conventional hydrometallurgical reprocessing, including reducing the reprocessing plant footprint, providing an initial degree of decontamination from fission products, and minimizing the amount of secondary waste from nitric acid. In this work, the direct extraction of uranium oxides into 1.5 M N,N-di(2-ethylhexyl)isobutyramide (DEHiBA) in n-dodecane is examined. UV–vis spectra and distribution ratios of HNO 2 in 1.5 M DEHiBA as well as the equilibrium organic phase H 2 O concentrations in HNO 3 -loaded 1.5 M DEHiBA are also reported. Hypothesized reaction stoichiometries for the direct extraction of uranium from UO 2 , α-U 3 O 8 , and ε-UO 3 are verified through analysis of organic-phase U, HNO 2 , and HNO 3 concentrations after dissolution. Water generated by the dissolution results in the formation of a separate aqueous phase, which will need to be accounted for in future flowsheet design.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

Analysis of trace elements in uranium by inductively coupled plasma-optical emission spectroscopy, design of experiments, and partial least squares regression

Inductively coupled plasma-optical emission spectra were modeled with partial least squares regression models to quantify trace elements in uranium (20–5000 μg per g U) without chemical separations or matrix matched calibration standards.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2026

The Role of Nuclear Data Sensitivities in Prompt α-Eigenvalue Predictions of Delayed Critical Benchmarks

Alpha (α) eigenvalues, which describe the logarithmic time derivative of the neutron population in a multiplying system, are integral to time-dependent behavior and diagnostic applications. However, uncertainties in the evaluated nuclear data can significantly impact the accuracy of transport simulations for such quantities. This work explores the use of machine learning models to predict two key outputs, α-eigenvalues and keff bias, using input features derived from α-eigenvalue sensitivities to nuclear data. The criticality safety benchmark models used in this study come from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Three models, random forest, XGBoost, and NGBoost, are trained on both energy-resolved and energy-summed α sensitivities. For the α-eigenvalue bias prediction, NGBoost achieved the highest R 2 (0.9476) using energy-resolved features, while XGBoost performed best using summed sensitivities. In contrast, when predicting the keff bias, all the models showed moderate predictive capability (best R 2 ≈ 0.72), as the mapping from the static α-sensitivities to the static keff bias was less direct. SHAP (SHapley Additive exPlanations) analysis was used to interpret the model predictions. Across both prediction tasks, the features associated with neutron capture [H-1 (n, γ)], uranium scattering reactions (such as 235 U elastic/inelastic), and actinide capture/fission reactions (such as 239 Pu and 234 U) were consistently identified as the most impactful. This highlights the key role of specific nuclear reactions and energy ranges in shaping both time-dependent and steady-state criticality behavior. These results demonstrated that α-sensitivities, despite being computed for time-dependent metrics, can provide valuable insights for predicting both α-eigenvalues and the keff bias. Moreover, machine learning models offer a promising pathway for uncovering important nuclear data dependencies and guiding future data evaluation efforts.

Nuclear data↗

DOE OSTI2020

AFIP6-MkII and RERTR-12 Porosity Data Collection and Analysis for Modeling and Simulation

Gathering data for the improvement of nuclear fuel modeling and simulation efforts is the primary driver for this work. Mechanistic models allow for a better understanding of the material on a micro- and macrostructural level while saving time and money over traditional experiment efforts. Historically, summarized data and correlations are the inputs for empirical material models and model validation. When improving these models for nuclear fuels with experimental results, there is a lack of reliable data readily available. Experiments - RERTR-12 and AFIP6-MkII - were conducted to understand the irradiation behavior of metallic U-10Mo monolithic fuels for use in extreme reactor environments such as research reactors like the Advanced Test Reactor (ATR) or the High Flux Isotope Reactor (HFIR). Microstructural characteristics of fission gas pores (FGP) in each experiment are collected using an automated image analysis technique developed at the University of Florida and presented here. A series of statistical tests are performed to explore the reliability of the results, as well as understand where the data is lacking and what future data collection is necessary to provide sufficient information to assist modeling efforts. The focus is on the porosity, pore size, and eccentricity of FGPs formed during irradiation in three AFIP6-MkII samples and one RERTR-12 sample. From the analysis, it is clear there are substantial impacts of fission density on the pore structure, but there also exist also underlying connections between each sample and the behavior observed in the pores. Further analyses of the pre- and post-irradiation microstructure are needed to improve the understanding of these connections. An early method for microstructural data analysis is presented within and is currently being expanded to include other microstructure data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Hydrogen Diffusion Coefficient Measures on Thin Film Uranium Oxide

Thin films of uranium oxides, putatively UO 2 and U 3 O 8 , were deposited on palladium/silver (75/25) foil discs (10 μm thick, 10 mm diameter). Methodology for sealing these foils, applying a hydrogen pressure (~ 1 atm) to the oxide side, and measuring the pressure on the permeate side (opposite side of foil) is reported. The experimental apparatus is held at 100 °C to speed diffusion, aided by the pressure differential across the foil (~1 atm (750 torr) on oxide side, initially low vacuum on permeate side, ~1x 10 -4 torr). Early results indicate an effective diffusion coefficient of roughly 7.66 x 10 -17 +/- 2.22 x 10 -17 cm 2 /sec for UO 2 . These values are in line with expectations and prior measures relative to large lag times for system baseline (substrate only). Evaluation of the technique for U 3 O 8 thin films suggests further development will be needed to extend the technique to more oxidized films (U 3 O 8 , UO 3 ). Characterization of the foils (thickness and speciation) post diffusion experiments will be carried out by SIGMA (Eric Tegtmeier and Andy Richards) in FY24 which will sharpen the uncertainty quantification for UO 2 diffusion coefficients, aiding the nucleation model for DRACO.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Credible Criticality Safety Margin in the 30B Package with LEU+ UF 6 and Hypothetical Water Ingress

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Development of Decay Heat Sensitivity Analysis Capability in SCALE/ORIGEN

In this work, a decay heat sensitivity analysis capability was developed and implemented in the ORIGEN code of the SCALE nuclear modeling and simulation suite. This capability introduces improved numerical integration schemes, which overcome the challenges associated with accurately modeling the behavior of adjoint nuclide amounts during coarse time steps for both nuclide amount and decay heat sensitivity calculations. This capability significantly improves the accuracy of calculations without compromising computational efficiency compared to the existing method. Extensive verification was conducted for various benchmark problems, including a 238 Pu decay and an irradiation problem involving 135 Xe, evaluated with both coarse and fine time grids. The results show excellent agreement with reference direct perturbation solutions, reaffirming the computational accuracy of the newly proposed numerical integration methods. Furthermore, sensitivity analyses were performed for fission product inventories ( 147 Sm, 150 Sm, 155 Gd) in pressurized water reactor UO 2 and MOX fuel assemblies. These analyses demonstrated that the ORIGEN sensitivity analysis capability can capture detailed sensitivity coefficients and underlying physics in real applications. Additionally, a decay heat sensitivity analysis for high-assay low-enriched uranium fuel, including various initial 235 U enrichment and burnup points, highlights the extended capabilities of SCALE/ORIGEN in comprehensively assessing the factors influencing total decay heat. These advancements in ORIGEN offer valuable insights for reactor analysis, fuel design, and safety assessments, especially in the context of advanced nuclear fuel development and design changes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Updated Guidance on Sensitivity/Uncertainty Methods Use for Nuclear Criticality Safety Validation

The use of sensitivity/uncertainty (S/U) methods in nuclear criticality safety (NCS) was established more than 25 years ago and has been increasing ever since. Recent interest in commercial applications involving uranium enrichments above 5 wt% 235 U has increased focus on demonstrating the applicability of existing critical experiments to this enrichment range. S/U techniques can also be particularly useful in assessing the similarity of new fuel forms in relation to those documented in existing published evaluations of benchmark critical experiments. Publishing updated recommendations for the use of these methods is thus important to ensure that new S/U practitioners can benefit from the years of development and application invested in generating a set of best practices to maximize the utility of these tools.

NCS↗

DOE OSTI2024

Viability of Miniature Discs in LFA-427

Miniature disc samples pose challenging measuring conditions for laser flash analysis. With reductions in signal from lower radiosity and lower half-rise times from low thicknesses, optimal selections of coatings and laser parameters are critical in a successful measurement. This report explains the methodology in successfully measuring two different material types—metals and ceramics—in proper coating selection, laser parameters, and post analysis for each material, as well as applying it to un-irradiated uranium samples of each type (U-metal, UO 2 ). This was done through a combination of analysis techniques combining automatic commercial methods (Neztech) and manual methods for correcting laser pulse impact aberrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗