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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 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Comparison of source-location algorithms for atmospheric samplers

Numerous algorithms have been developed to determine the source characteristics for an atmospheric release, e.g., (Bieringer et al., 2017). This study compares three models that have been applied to the data collected by the International Monitoring System operated by the Comprehensive Nuclear-Test-Ban Treaty Preparatory Commission to estimate source event parameters. Each model uses a different approach to estimate the parameters. A deterministic model uses a possible source region (PSR) approach (Ringbom et al., 2014) that is based on the correlation between predicted and measured sample values. An unnamed model developed at Pacific Northwest National Laboratory (herein called BAYEST) uses a Bayesian formulation (Eslinger et al., 2019; Eslinger et al., 2020; Eslinger and Schrom, 2016). The FREAR model uses a different Bayesian formulation (De Meutter and Hoffman, 2020; De Meutter et al., 2021; Pieter De Meutter, 2021). The performance of the three source-location models is evaluated with 100 synthetic release cases for the single xenon isotope, 133 Xe. The release cases resulted in detections in a fictitious network with 120 noble gas samplers. All three source-location models use the same sampling data. The two Bayesian models yield more accurate location estimates than the deterministic PSR model, with FREAR having slightly better location performance than BAYEST. Samplers with collection periods of 3, 6, 8, 12, and 24-h were used. Results from BAYEST show that location accuracy improves with each reduction in sample collection length. The BAYEST model is slightly better for estimating the start time of the release. The PSR model has about the same spread in start times as the FREAR model, but the PSR results have a better average start time. The Bayesian source-location algorithms give more accurate results than the PSR approach, and provide release magnitude estimates, while the base PSR model does not estimate the release magnitude. This investigation demonstrates that a reasonably dense sampling grid will sometimes yield poor location and time estimates regardless of the model. The poor estimates generally coincide with cases where there is a much larger distance between the release point and the first detecting sampler than the average sampler spacing.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Deployment of portable, modular gas samplers as part of an atmospheric tracer experiment

Underground nuclear explosions release noble gases into the atmosphere that can be detected to support international monitoring efforts. Atmospheric transport models help predict the movement of these gases over long distances, but struggle to predict the movement in the atmosphere local to the release. A field experiment was designed to monitor the movement of 127 Xe within a 5-km radius. Four gas samplers were deployed as part of this experiment to collect atmospheric samples at various distances from the release point. In conclusion, these samples were then analyzed in a near-field lab using a NaI detector and in an off-site lab using gamma-gamma coincidence and beta-gamma coincidence counting.

Radioxenon↗
DOE OSTI2020

Protection of graphite from salt and gas permeation in molten salt reactors

The reactor core, moderator and reflectors of a thermal spectrum advanced molten salt reactor will constitute multi-tons of graphite. Porous reactor-grade graphite, if unprotected, can be permeated by molten salt depending on the infiltrating pressure differential and entrance diameters of accessible graphite pores. Salt and gas permeation of graphite can affect microstructural properties and radiation behavior but also facilitate diffusion, deposition and retention of fission products and tritium. Because of the significant void volume of nuclear graphite, fission products and tritium retention due to salt permeation necessitates seal coatings or pore impregnation to reduce open porosity. Alternatively, very fine-grained graphite grades with low Xe permeability are being developed. In this work, we survey the current technologies for mitigating salt and gas transport into graphite.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Radiation driven diffusion in γU-Mo

A monolithic fuel design based on a U-Mo alloy has been selected as the fuel type for conversion of the United States High-Performance Research Reactors (HPRRs). A critical phenomenon of interest with U-Mo monolithic fuel is the large amount of swelling that takes place during operation, particularly at high fission densities. The accurate prediction of fuel evolution under irradiation requires implementation of correct thermodynamic and kinetic properties into mesoscale and continuum level fuel performance modeling codes. One such property where there exists incomplete data is the diffusion of relevant species under irradiation. Fuel performance swelling predictions rely on an accurate representation of diffusion in order to determine the rate of fission gas swelling and the local microstructural evolution. In this work, we present molecular dynamics simulations of the radiation driven diffusion of U, Mo and Xe in U-Mo nuclear fuels. Diffusion coefficients for each species are determined over a range of temperatures and compositions. In this work, updated diffusion coefficients are presented that are applicable under irradiation that incorporate both intrinsic and radiation driven diffusion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Temperature Effects on Interdiffusion of Al and U-Mo under Irradiation

A high-energy Xe ion irradiation experiment was conducted to investigate the temperature dependence of interdiffusion in bilayer Al-UMo samples under irradiation. The amount of interdiffusion achieved at a fixed dose with the increase of temperature showed a clear transition at 175 degrees C (with an estimated error in the range of ± 10 degrees C) from temperature-independent to temperature-dependent behavior. The activation energy derived from the curve of interdiffusion quantity vs. irradiation temperature is 0.77 ± 0.16 eV. This information has been utilized to understand the temperature effect on the interdiffusion process that occurred at the interfaces of U-Mo particles and the Al matrix in U-Mo/Al dispersion fuels, whose magnitude significantly impacts the fuel's performance. Although this temperature effect was deemed important, it cannot be examined directly using in-pile irradiation data, as fuel temperatures cannot be measured in reactor irradiation and are highly correlated with fission rate and thermal conductivity evolution. To connect the knowledge accumulated from ion irradiation with in-pile irradiation data, simulation of a full-sized U-Mo/Al dispersion fuel plate irradiated in the FUTURE test in the BR2 reactor was performed with the Dispersion Analysis Research Tool (DART), a dispersion fuel performance code. DART is equipped with an interaction or interdiffusion layer (IL) growth correlation formulated to describe the temperature dependence of ion mixing results. The agreement between calculated and measured fuel meat constituent volume fractions and swelling data demonstrated that the temperature effect on in-pile Al-UMo interdiffusion is well captured with the correlation. In this case, the fitted activation energy is 0.70 eV. Considering the uncertainties associated with the ion irradiation data, the activation energy obtained from in-pile data fitting is in accord with that from ion irradiation results.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Depth-resolved thermal conductivity and damage in swift heavy ion irradiated metal oxides

We investigated thermal transport in swift heavy ion (SHI) irradiated insulating single crystalline oxide materials: yttrium aluminum garnet- Y 3 Al 5 O 12 (YAG), sapphire (Al 2 O 3 ), zinc oxide (ZnO) and magnesium oxide (MgO) irradiated by 167 MeV Xe ions at 10 12 – 10 14 ions/cm 2 fluences. Depth profiling of the thermal transport on nano- and micro- meter scales was assessed by time-domain thermoreflectance (TDTR) and modulated thermoreflectance (MTR) methods, respectively. This combination allowed us to isolate the conductivities of different sub-surface damage-regions characterized by their distinct microstructure evolution regimes. Thermal conductivity degradation in SHI irradiated YAG and Al 2 O 3 is attributed to formation of ion tracks and subsequent amorphization, while in ZnO and MgO it is mostly due to point defects. Additionally, notably lower conductivity when probed by very low penetrating thermal waves is consistent with surface hillock formation. An analytical model based on Klemens-Callaway method for thermal conductivity coupled with a simplified microstructure evolution capturing saturation in defect concentration was used to obtain depth dependent damage across the ion impacted region. The studies showed that YAG has the highest damage profile resulting in the less dependence of thermal conductivity with the depth, while MgO on the contrary has the strongest dependence. The presented work sheds new light on how SHI induced defects affect thermal transport degradation and recovery of oxide ceramics as promising candidates for next generation nuclear reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Measurement of Helium and Xenon Knudsen permeability in graphite IG-11

Here, understanding the flow and diffusion (transport) of gases through nuclear graphite is of interest in both the prismatic and pebble bed high-temperature gas reactors (HTGRs), for normal operation and under accident conditions, for example, as related to graphite oxidation. Both the laminar and turbulent gaseous transport are of interest. It has also been noted that nuclear graphite can have pore sizes that can possibly lead to non-continuum transport, as the pore size can be comparable to the gaseous mean free path. We report measurements and analyzes of experiments performed with IG-11 graphite in the pressure range of 10 Pa to 14,700 Pa, 293 K, using Helium, and research grade xenon (99.999%), respectively. These values are in fair agreement with those reported for different graphites and gases by other investigators, but are also higher by six orders of magnitude than those reported for Kr permeability in a developmental very low permeability graphite-HXT-90. Our measured ratio of the Helium and Xenon permeabilities is about 4, different from $\sqrt{\frac{m_{Xe}}{m_{He}}}$ ≈ 5.72 that corresponds to diffuse reflection at the graphite surface for both gases. Simple analysis indicates that the two gases have slightly different accommodation coefficients (about 25%) with the graphite surface. However, there could also be other reasons, such as experimental uncertainties or the need for a more detailed analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2022

An investigation of the failure modes in U-10Mo monolithic fuel irradiated to high burnup

Here, this study investigated possible failure mode sequences in U-10Mo monolithic fuel irradiated to very-high-burnup. The U-Mo fuel plate used in this study was characterized using scanning electron microscopy (SEM), wavelength dispersive spectroscopy (WDS), and image analysis techniques to investigate how microstructural features may initiate crack formation and propagation through the microstructure, potentially leading to blistering. Distinctly large fission gas pores (FGPs) were observed to preferentially align 5-15µm away from the U-10Mo/Zr interface, parallel to the Zr diffusion barrier. Preferential growth, alignment, and interconnection of FGP could initiate blistering in the fuel plate by creating a fission gas channel, parallel to the U-Mo/Zr interface. FGP alignment in the fuel phase near the U-Mo/Zr interface is believed to be one of the precursors to crack and Type 2 blister formation in monolithic U-Mo fuels. The chemical maps revealed differences in fission product behavior near the U-Mo/Zr interface such that Nd remains in the fuel matrix at the FGP sites. On the other hand, other fission products like Xe and Cs can get trapped in FGPs by Nd in the fuel phase and/or diffuse toward the Zr diffusion barrier.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermal conductivity degradation due to radiation-induced amorphization in U 3 Si 2 : A pilot study

Here, in this study, we investigate the thermal conductivity of U 3 Si 2 amorphized by ion irradiation using 84 MeV 136 Xe ions at 190 °C. The suspended-bridge method was utilized to measure the thermal conductivity, allowing for a detailed analysis of the specimen while minimizing interference from other crystalline phases. Our results indicate that the thermal conductivity of amorphous U 3 Si 2 is significantly lower than that of unirradiated crystalline U 3 Si 2 . These findings are consistent with recent studies on in-pile-irradiated U 3 Si 2 samples that consider the effects of U 3 Si 2 amorphization, fission gas bubbles, and other impurities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Synchrotron micro-computed tomography analysis of neutron-irradiated U-Mo fuel

The three-dimensional (3D) microstructure of neutron-irradiated uranium-10 wt.% molybdenum (U-10Mo) fuel with a burn-up of 9.8 × 10 21 fissions/cm 3 was characterized using a novel, multi-modal synchrotron micro-computed tomography approach combining propagation-based phase-contrast enhanced and absorption contrast techniques. The porosity development, porosity interconnectedness, swelling, composition, local thickness of the zirconium (Zr) diffusion barrier, and the influence of the fuel–cladding interaction on the local composition and pore morphology, were uniquely determined in 3D. Two cuboids were produced using a focused ion beam-scanning electron microscope at the Zr diffusion barrier–fuel interface and in the bulk fuel. The bulk fuel sample swelled by 53.3 [+9.7/−3.1]%, while the fuel near the Zr–fuel interface swelled by 63.3 [+14.7/−7.2]%. The average local thickness of the Zr diffusion barrier decreased by 53 %, compared to the expected pre-irradiated thickness. Four pore morphology regions were identified initiating parallel to the fuel–Zr interaction region: (1) an interaction layer of suppressed porosity, (2) a layer of elongated and interconnected porosity, (3) a transition zone of low porosity, and (4) a layer of unoriented porosity representative of the bulk fuel behavior. The increase in porosity near the diffusion barrier corresponded to a higher U concentration compared to that in the bulk fuel. The interconnected porosity in the fuel near the diffusion barrier was extensive and oriented parallel to the diffusion barrier, while the bulk fuel had more compact and isolated pore networks. The interaction layer, despite having suppressed porosity, was nearly 100 wt.% U. Porosity suppression at the diffusion barrier corresponds to the expected reduction in radiation-driven diffusion of Xe at the interface despite the anticipated increase in fission product nucleation originating from a higher U concentration. In conclusion, the novel 3D insights of the porosity, swelling, and compositional variations characterized herein can improve the fidelity of fuel performance codes for proliferation-resistant fuels for research and test reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

In-situ irradiation of uranium carbide

Uranium carbide (UC) is a leading candidate fuel for Generation IV reactors due to its high uranium density and thermal conductivity. However, its irradiation performance—particularly gas bubble swelling and defect dynamics—remains poorly characterized. Using in-situ transmission electron microscopy (TEM), we irradiated UC with 300 keV Xe + and 1 MeV Kr 2+ ions at temperatures up to 900 °C to quantify swelling behavior and dislocation loop evolution. The swelling remained below 0.6 % across all temperatures, suggesting the dimensional stability of UC under irradiation at these temperatures. Dislocation loops grew faster in UC than in UO 2 or UN, correlating with its lower homologous temperature. Notably, nanograin structures emerged in thin regions of the lamellar, mirroring phenomena previously observed in UO 2 and ZrC. These results address critical knowledge gaps in the radiation tolerance of UC and provide insight into its suitability for advanced reactor systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Effect of TiN coating on suppressing Ce-Fe interaction under irradiation

Advanced cladding is critical for fast reactors with the adequate thermal conductivity, mechanical stability and radiation tolerance of the cladding base material, corrosion resistance and high temperature coolant compatibility of the cladding surface, and chemical stability of the cladding inner wall against fuel cladding chemical interaction (FCCI). The preliminary results of recent ion irradiation studies of two diffusion-couple samples of cerium (Ce)/oxide-dispersion strengthened steel (ODS) and Ce/TiN/ODS, irradiated with 80 MeV xenon (Xe) ions to 100 displacements per atom (dpa) at 500°C, are summarized. Significant Ce-Fe interaction occurred in the Ce/ODS sample, and no noticeable Ce-Fe interaction was found in the Ce/TiN/ODS sample. It shows the effectiveness of 1-µm TiN diffusion barrier coated by the pulsed laser deposition on suppressing Ce-Fe interaction, a major contributor to FCCI in cladding. Here, density function theory (DFT) calculations of the impurity diffusivities of Ce and Fe within the Ti sublattice of TiN were performed to assist a mechanistic understanding of the experimental results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Nuclear waste attributes of near-term deployable small modular reactors

The nuclear waste attributes of near-term deployable SMRs were assessed using established nuclear waste metrics, which are the DU mass, SNF mass, volume, activity, decay heat, radiotoxicity, and decommissioning LLW volumes. Metrics normalized per unit electricity generation were compared to a reference large PWR. Three SMRs, VOYGR, Natrium, and Xe-100, were selected because they represent a range of reactor and fuel technologies and are active designs deployable by the decade’s end. The SMR nuclear waste attributes show both some similarities to the PWR and some significant differences caused by reactor-specific design features. The DU mass is equivalent to or slightly higher than the PWR. Back-end waste attributes for SNF disposition vary, but the differences have a limited impact on long-term repository isolation. SMR designs can vary significantly in SNF volume (and thus heat generation density). However, these differences are amenable to design optimization for handling, storage, transportation, and disposal technologies. Nuclear waste attributes from decommissioning vary depending on design and decommissioning technology choices. Given the analysis results in this study and assuming appropriate waste management system and operational optimization, there appear to be no major challenges to managing SMR nuclear wastes compared to the reference PWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The LUX-ZEPLIN (LZ) Experiment

We describe the design and assembly of the LUX-ZEPLIN experiment, a direct detection search for cosmic WIMP dark matter particles. The centerpiece of the experiment is a large liquid xenon time projection chamber sensitive to low energy nuclear recoils. Rejection of backgrounds is enhanced by a Xe skin veto detector and by a liquid scintillator Outer Detector loaded with gadolinium for efficient neutron capture and tagging. LZ is located in the Davis Cavern at the 4850’ level of the Sanford Underground Research Facility in Lead, South Dakota, USA. We describe the major subsystems of the experiment and its key design features and requirements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Differences in the response of two light guide technologies and two readout technologies after an exchange of liquid argon in the dewar

In this investigation the response to the scintillation light generated by through-going cosmic muons in liquid argon (LAr) was measured by two light guide technologies and two readout technologies after five weeks of running in the TallBo dewar at Fermilab. The response was remeasured after the dewar was drained of LAr, refilled, and then run again for an additional four weeks. After the dewar was refilled, there was clear evidence that the scintillation signal had dropped significantly. The two light guide technologies were developed at Indiana University and MIT/Fermilab. The two readout technologies were boards that passively or actively ganged 12 Hamamatsu MPPCs. Two possible explanations were identified for the degraded signal: the response of the two light guide technologies degraded due to damage caused by thermal cycling, and/or unknown differences in the trace residual Xe contamination in the fills of LAr led to the observed drop in scintillation light. Neither absorption nor quenching by N 2 , O 2 , and H 2 O contamination can account for the degradation. Neither the individual Hamamatsu MPPCs nor the passive/active ganging boards appear to have been affected by the thermal cycling. The path length distributions of the cosmics traversing the dewar appear quite similar in both event samples.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Event reconstruction in a liquid xenon Time Projection Chamber with an optically-open field cage

nEXO is a proposed tonne-scale neutrinoless double beta decay (0νββ) experiment using liquid 136 Xe (LXe) in a Time Projection Chamber (TPC) to read out ionization and scintillation signals. Between the field cage and the LXe vessel, a layer of LXe (“skin” LXe) is present, where no ionization signal is collected. Only scintillation photons are detected, owing to the lack of optical barrier around the field cage. Here, in this work, we show that the light originating in the skin LXe region can be used to improve background discrimination by 5% over previous published estimates. This improvement comes from two elements. First, a fraction of the γ-ray background is removed by identifying light from interactions with an energy deposition in the skin LXe. Then, background from 222 Rn dissolved in the skin LXe can be efficiently rejected by tagging the α decay in the 214 Bi- 214 Po chain in the skin LXe.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Grain subdivision and structural modifications by high-energy heavy ions in UO 2 with different initial grain size

Irradiation with 84 MeV Xe ions at peak fluence above 7 × 10 17 ions/cm 2 has been carried out on two UO 2 samples with different initial grain sizes at 300 °C and 500 °C. Characterization of both grain structure and bubble morphology has been performed using transmission electron microscopy, image analysis and precession electron diffraction. The bubble size in the large-grained sample was smaller than that in the standard-grain sample, confirming fission gas precipitation resistance in the former. The ion irradiation in the high electronic energy deposition range has caused formation of a large density of dislocations which resulted in grain polygonization for the sample with standard grain size, which was observed in multiple lamellas extracted next to each other at peak dose. The sample with large grains showed inconsistent behavior. In one case, dislocation pile-up occurred, but no sub-grain formation was observed. An adjacent location showed sub-grain formation, even if to less extent compared to the standard grain-size sample. Furthermore, these findings suggest a limited polygonization resistance in large-grained UO 2 compared to standard-grain UO 2 . Recrystallization of both sample surfaces was present, leaving unclear if this is a surface effect or a consequence of the highest energy deposition rate in this area.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Characterizing the initial state and dynamical evolution in XeXe and PbPb collisions using multiparticle cumulants

For the first time, correlations among mixed-order moments of two or three flow harmonics —( v n k , v m l ) and ( v n k , v m l , v p q ), with k, l, and q denoting the respective orders—are measured in xenon–xenon (XeXe) collisions and compared with lead–lead (PbPb) results, providing a novel probe of collective behavior in heavy ion collisions. These measurements compare a nearly spherical, doubly-magic 208 Pb nucleus to a triaxially deformed 129 Xe nucleus, emphasizing the sensitivity to initial-state geometry fluctuations arising from nuclear deformation. The dependence of these results (v n , n = 2 , 3 , 4 ) on the shape and size of the nuclear overlap region is studied. Comparisons between v 2 , v 3 , and v 4 demonstrate the importance of v 3 and v 4 in exploring the nonlinear hydrodynamic response of the quark-gluon plasma (QGP) to the initial spatial anisotropy. The results constrain initial-state model parameters that influence the evolution of the QGP. The CMS detector was used to collect XeXe and PbPb data at nucleon-nucleon center-of-mass energies of S N N = 5.44 and 5.36 TeV, respectively. Correlations are extracted using multiparticle mixed-harmonic cumulants (up to eight-particle cumulants) with charged particles in the pseudorapidity range |η| < 2.4 and transverse momentum range 0.5 < p T < 3 GeV/c.

CMS↗