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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Length scale effects of micro- and meso-scale tensile tests of unirradiated and irradiated Zircaloy-4 cladding

Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso-scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Here, meso-scale testing found strength and strain values similar to that of bulk-scale testing.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Estimation of Fission Product Transport Parameters for Cesium in the AGR-3/4 TRISO Fuel Experiment

A one-dimensional (1D) finite-element model of fission product transport in the AGR-3/4 experiment has been developed using the Multiphysics Object Oriented Simulation Environment (MOOSE) framework and implemented in the fuel performance code, BISON. The model resolves capsule-specific geometries, materials, and temperature histories and simulates radial migration of fission products from the fuel compact through the inner ring, outer ring, and into the sink ring. Model parameters governing diffusion and sorption were estimated for key fission products – cesium (Cs), and europium (Eu) – by simultaneously fitting modeled isotopic concentration profiles and total ring inventories to a post-irradiation experimental measurement. These data include gamma scanning, liquid scintillation for Sr-90, radial deconsolidation leach-burn-leach analysis, tomographic reconstructions, and destructive physical sampling. A mortar-based interfacial sorption framework was implemented to enforce physically consistent mass transfer and flux conservation across gas gaps. Two classes of parameter sets were derived: a least-squares best-fit, and a safety-oriented conservative-fit, what applies strong penalties for underprediction of sink inventories. Across all twelve capsules, the model successfully reproduces the dominant radial transport trends for Cs, Sr, with decreasing concentrations from the compact outward through successive rings. Cs behavior is captured most consistently, while strontium predictions reveal systematic trade-offs between compact accuracy and conservative sink-ring bounding. The results demonstrate that sink ring weighted calibration provides conservative, safety relevant bounds on low temperature fission product transport, but at the cost of underpredicting compact inventories for Sr isotopes. These discrepancies highlight the need for additional physics, including fast-slow diffusion model, incorporating trapping mechanism in the transport behavior. Overall, this work establishes a robust, capsule-specific modeling framework for AGR-3/4 fission product transport and provides a defensible basis for parameter selection in source-term and fuel performance analyses for high temperature gas-cooled reactors.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Improvement of the $\mathrm{BISON U_3Si_2}$ modeling capabilities based on multiscale developments to modeling fission gas behavior

Uranium silicide (U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel, but operated at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. These lower length scale informed models are then utilized in the validation of BISON U 3 Si 2 modeling capabilities to simulate the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the validation process to identify where further experiments and lower length scale modeling would be beneficial. Here, the multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Chemical characterisation of degraded nuclear fuel analogues simulating the Fukushima Daiichi nuclear accident

The Fukushima Daiichi accident generated degraded nuclear fuel material, mixed with other reactor components, known as molten core-concrete interaction (MCCI) material. Simulant MCCI material was synthesised, excluding highly radioactive fission products, containing depleted U, and incorporating Ce as a surrogate for Pu. Multi-modal µ-focus X-ray analysis revealed the presence of the expected suite of U-Zr-O containing minerals, in addition to crystalline silicate phases CaSiO 3 , SiO 2 -cristobalite and Ce-bearing percleveite, (Ce,Nd) 2 Si 2 O 7 . The formation of perclevite resulted from reaction between the U-Zr-O-depleted Ce-Nd-O melt and the silicate (SiO 2 ) melt. It was determined that the majority of U was present as U 4 , whereas Ce was observed to be present as Ce 3+ , consistent with the highly reducing synthesis conditions. A range of Fe-containing phases characterised by different average oxidation states were identified, and it is hypothesised that their formation induced heterogeneity in the local oxygen potential, influencing the oxidation state of Ce.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Acoustic sensing and autoencoder approach for abnormal gas detection in a spent nuclear fuel canister mock-up

Currently, spent nuclear fuel (SNF) from commercial nuclear power plants is stored in stainless-steel canisters for interim dry storage. To provide an inert environment, these canisters are backfilled with helium after vacuum drying. However, the helium environment may be contaminated during extended storage because of the material degradation. For example, the heavier fission gas xenon may be released from the fuel rods into the canister cavity should the fuel cladding be breached. Other gases such as air and water vapor may also be present as a result of leakage caused by chloride-induced stress corrosion cracking on the canister walls or by insufficient vacuum drying. Therefore, monitoring the gas composition can provide critical information about the health of SNF canisters. In this study, noninvasive testing was conducted on a 2/3-scaled SNF canister mock-up using acoustic sensing. Ultrasonic transducers were placed on the exterior surface of the canister to probe the gas composition. A dataset was collected by sealing the canister mock-up and introducing up to 1.53% argon or 1.29% air into the helium background gas. Three methods were used to detect changes in the gas composition: the time-of-flight (TOF) method, the differential method, and the autoencoder method. Results showed that the TOF method had sufficient resolution to detect abnormal gas concentrations of less than 1.0%. The differential method demonstrated a periodic in-phase and out-of-phase behavior between the benchmark (i.e., pure helium) and abnormal (i.e., with argon or air) state signals. The variational autoencoder (VAE) and the Wasserstein autoencoder (WAE) were trained on the benchmark data and were applied directly to the abnormal state data. It was found that both the unsupervised VAE and the WAE were able to distinguish the benchmark and abnormal states of the canister mock-up based on the reconstruction error.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Temperature-dependent cavity swelling in dual-ion irradiated Fe and Fe-Cr ferritic alloys

Fe-Cr ferritic-martensitic (FM) steels are promising structural material candidates for fusion and advanced fission reactors due to their attractive mechanical properties and volumetric swelling resistance. However, significant discrepancies exist regarding the effect of solutes and irradiation temperature on cavity swelling under ion versus neutron irradiation conditions. In this study, simultaneous dual ion irradiations (8 MeV Ni 3+ ions and energy-degraded 3.5 MeV He 2+ ions) were used to quantify the cavity swelling behavior in ultra-high purity Fe and Fe-Cr alloys (3-14 wt.% Cr), Fe-10 wt.% Cr-780 wt.ppm C, and Eurofer97 FM steel. The irradiations were conducted over a wide temperature range (400-550°C) with a mid-range dose of ~30 displacements per atom (dpa) and 0.1 appm/dpa He implantation rate. Here, using state-of-the-art transmission electron microscopy (TEM), we reveal that pure Fe has a ~50°C lower peak swelling temperature difference than Fe-Cr alloys, which is attributed to higher vacancy mobility in pure Fe. Chromium solute appears to strongly suppress cavity swelling in Fe-Cr alloys for temperatures below ~470°C, but seems to have little effect or slightly enhances swelling above ~470°C. Cavities were observed in all the irradiated samples between 400-550°C. This indicates that the narrow temperature range of observable cavities reported in prior ion irradiated Fe-Cr ferritic alloy studies is likely an artifact associated with the use of low ion energies (<5 MeV), which leads to pronounced near-surface and implanted ion effects that suppress cavity swelling even at midrange depths (particularly at high temperatures).

36 MATERIALS SCIENCE↗
DOE OSTI2026

On the uncertainties in helium generation predictions for fission and fusion alloys

With ongoing advances in fusion and advanced fission reactors, quantifying irradiation effects in materials is critical. Transmutation-induced helium in cladding and structural materials can drive swelling and embrittlement, thereby reducing these components’ lifespans. Yet most studies ignore the considerable uncertainties in predicting helium generation rates. In this work, we created a code wrapper, F-SCATTER, that automatically performs simulations in FISPACT-II. We used this tool to investigate potential variance in helium generation rate, or He/dpa, calculations based on deviations in alloy composition, irradiating neutron flux spectrum, computational methodology, and nuclear data sources. We used 12 wt% Cr HT9 steel as the reference case and observed a 6.5%–98.3% He/dpa spread based on compositional variation within a single chemical specification, a 1.8%–11.5% He/dpa variation upon the incorporation of a 15% artificial uncertainty in flux at each energy, and a He/dpa difference as high as 231% when using ENDF/B-VIII.0 versus TENDL-2021 data libraries. Similar results were found for other prominent iron-based alloys, including Grade 91, castable nano-structured alloy, and 316H—where additional variations exist based on reactor type (e.g. thermal, fast, or fusion) and alloying elements such as carbon, nitrogen, and nickel. Based on the simulated results, we conclude that a significant part of the heat-to-heat variability in swelling responses of Fe-based alloys can be driven by impurity content in alloy compositions, and, therefore, chemical control should be a key element in supply chain design for advanced nuclear energy systems. Furthermore, we provide critical recommendations on best practices for evaluating and reporting helium production and lattice damage rates when computing predictions with multiphysics programs such as FISPACT-II.

FISPACT-II↗
DOE OSTI2020

Inline gamma-spectrometry of fission product elements after rapid high-pressure ion chromatographic separation

Analysis of irradiated material shortly after irradiation can be non-trivial due to highly radioactive activation and fission isotopes increasing dead time in gamma-ray detection systems, often requiring a “cooling-period” between receipt of a sample and the subsequent analysis. A direct separation–detection method has been developed combining ion chromatography and inductively coupled mass spectrometry for rapid, low-level analysis of fission products; it cannot, however, detect certain short-lived species below the detection limit of the system. In this paper we report the implementation of an inline gamma-ray detector, which was added post-separation, pre-analysis, to test the quality and utility of elementally-isolated gamma-ray spectroscopy.

Inline gamma-ray detection↗
DOE OSTI2025

Multiscale, mechanistic modeling of cesium transport in silicon carbide for TRISO fuel performance prediction

Understanding cesium (Cs) transport in TRistructural ISOtropic (TRISO) particle fuel is crucial for predicting fission product release in high-temperature reactors. However, current challenges include significant scatter in diffusivity data and unexplained temperature-dependent diffusion regimes in the silicon carbide layer. This study addresses these challenges by developing a multiscale, mechanistic Cs transport model integrating atomistic simulations and phase field modeling. Our model quantifies temperature and grain size effects on Cs diffusivity, attributing experimentally observed regimes to a transition from bulk-dominated diffusivity at high temperatures to grain boundary-dominated diffusivity at lower temperatures. The model, validated against diffusion measurements and advanced gas reactor (AGR)-1 and AGR-2 post-irradiation fission product release data, enhances the predictive capability of the BISON fuel performance code. This study advances our understanding of Cs release from TRISO particles and its dependence on temperature and silicon carbide grain size, with implications for the safety and efficiency of high-temperature nuclear reactors.

BISON↗
DOE OSTI2023

An efficient instance segmentation approach for studying fission gas bubbles in irradiated metallic nuclear fuel

Abstract Gaseous fission products from nuclear fission reactions tend to form fission gas bubbles of various shapes and sizes inside nuclear fuel. The behavior of fission gas bubbles dictates nuclear fuel performances, such as fission gas release, grain growth, swelling, and fuel cladding mechanical interaction. Although mechanical understanding of the overall evolution behavior of fission gas bubbles is well known, lacking the quantitative data and high-level correlation between burnup/temperature and microstructure evolution blocks the development of predictive models and reduces the possibility of accelerating the qualification for new fuel forms. Historical characterization of fission gas bubbles in irradiated nuclear fuel relied on a simple threshold method working on low-resolution optical microscopy images. Advanced characterization of fission gas bubbles using scanning electron microscopic images reveals unprecedented details and extensive morphological data, which strains the effectiveness of conventional methods. This paper proposes a hybrid framework, based on digital image processing and deep learning models, to efficiently detect and classify fission gas bubbles from scanning electron microscopic images. The developed bubble annotation tool used a multitask deep learning network that integrates U-Net and ResNet to accomplish instance-level bubble segmentation. With limited annotated data, the model achieves a recall ratio of more than 90%, a leap forward compared to the threshold method. The model has the capability to identify fission gas bubbles with and without lanthanides to better understand the movement of lanthanide fission products and fuel cladding chemical interaction. Lastly, the deep learning model is versatile and applicable to the micro-structure segmentation of similar materials.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Colloquium : Neutrino detectors as tools for nuclear security

For over 40 years, physicists have discussed possible uses for neutrino detectors in nuclear nonproliferation, arms control, and fissile materials security. Neutrinos are an attractive fission signature because they readily pass through matter. The same property makes neutrinos challenging to detect in systems that would be practical for nuclear security applications. This Colloquium presents a broad overview of several potential neutrino applications, including the near-field monitoring of known reactors, far-field monitoring of known or discovery of undeclared reactors, detection of reactor waste streams, and detection of nuclear explosions. Recent detector advances have made near-field monitoring feasible, whereas farther-field reactor detection and waste stream detection monitoring may be possible in some cases with further research and development. Very long-range reactor monitoring and nuclear explosion detection do not seem feasible for the foreseeable future due to considerable physical and/or practical constraints.

42 ENGINEERING↗
NASA NTRS1972

Examination of UC-ZrC after long term irradiation at thermionic temperature

Two fluoride tungsten clad UC-ZrC fueled capsules, designated as V-2C and V-2D, were examined a hot cell after irradiation in NASA Plum Brook Reactor at a maximum cladding temperature of 1930 K for 11,089 and 12,031 hours to burnups of 3.0 x 10 to the 20th power and 2.1 x 10 to the 20th power fission/c.c. respectively. Percentage of fission gas release from the fuel material was measured by radiochemical means. Cladding deformation, fuel-cladding interaction and microstructures of fuel, cladding, and fuel-cladding interface were studied metallographically. Compositions of dispersions in fuel, fuel matrix and fuel-cladding interaction layer were analyzed by electron microprobe techniques. Axial and radial distributions of burnup were determined by gamma-scan, autoradiography and isotopic burnup analysis. The results are presented and discussed in conjunction with the requirements of thermionic fuel elements for space power application.

Yang, L.↗
NASA NTRS1984

The origin of the Moon

Bulk density alone shows that the Moon is depleted in metallic FeNi relative to the Earth or to chondritic meteorites. This depletion implies that the Moon formed not from chondrites but from differentiated material. Origin of the Moon by fission from the Earth offers a simple explanation for its depletion in FeNi, but this mechanism seems unlikely because of associated dynamical difficulties. Lunar volatile element depletions were invoked in support of fission, but volatile contents of eucritic meteorites are similarly low and the eucrites did not form by Earth fission. A more plausible origin of the Moon is accretion from the circumterrestrial swarm. The low FeNi content of the Moon is understood if the mean size of interplanetary silicate particles was much smaller than that for metal particles, since this would have led to preferential capture of silicates into Earth orbit, but the question arises whether the mean particle size of the metallic particles was great enough to prevent their capture into the swarm.

Wasson, J. T.↗
DOE OSTI2020

Sulfonamide and Sulfonamido-phenol Ligands for Extraction of f-Elements from Alkaline High-Level Waste

Alkaline High Level Waste (HLW) has been accumulated at Hanford and Savannah River Sites as a result of reprocessing for nuclear weapons production during the cold war. A large volume (∼100 MGal) has been accumulated in carbon steel tanks at Savannah River (SRS) and Hanford. The tank waste contains three separate phases generated when NaOH was added to previously acidic Purex raffinates: 1-2 Supernatant liquid, salt-cake, and sludge. The sludge consists mainly of insoluble hydroxides of transition metals while the supernate and salt-cake contain caustic-soluble materials, including salts of highly radioactive fission products Cs(I) and Sr(II). Current treatment of alkaline HLW in SRS includes: i) The Actinide Removal Process (ARP), which is based on sorption of {sup 90}Sr and Actinides (An) on monosodium titanate (MST), also known as 'alpha-strike' process, followed by ii) Caustic Side Solvent Extraction (CSSX),4 which is used for the extraction of {sup 137}Cs by modified calixarenes in a hydrocarbon diluent. Residual actinides in some tanks are removed after CSSX by an additional ARP process commonly referred to as 'alpha-finishing'. Despite the success of ARP for Sr and An removal, as it is a sorption process, it represents the kinetic bottleneck of integrated salt waste processing. Hence potential introduction of additional organic ligands for actinide extraction (in a modified CSSX process) could simplify the overall integrated process, making it more efficient and economical, with less titanate needed and shorter sorption time, as some of the actinide component would be removed during CSSX. In this study tri-sulfonamide and o-sulfonamido-phenol (mono-sulfonamide) ligands have been studied as extractants for Sm(III), which is being used as an Am(III) surrogate. Our prior studies in the group using a tri-sulfonamide (iPr-tsa-B) showed favorable extraction for Sm(III) nitrate salts from alkaline solutions. Mono-sulfonamides possess similar orientation of binding sites to pyrocatechols, which have been found to be good ligands for Am(III) binding and extraction from alkaline media. Tri-sulfonamide of the type iPr-tsa-B6 (1 mM in CH{sub 2}Cl{sub 2} solution) was studied for Ln{sup 3+} extraction using Sm(NO{sub 3}){sub 3}.6H{sub 2}0 (10 and 25 μM) in alkaline solution of NaOH (0.05, 0.1, 0.2, 0.3 mM) / 0.1 M NaNO{sub 3}. Stripping of the organic phase was done using 0.1 M HNO{sub 3} and quantification of Sm{sup 3+} was done using ICP-OES at 359.3 nm. The need to improve stability of the complex led to synthesis of compounds with N-donor site closer to the central benzene ring to facilitate cation-π interactions. Synthesis of tri-sulfonamide type A: a) Chloromethyl methyl ether, SnCl{sub 4}, CH{sub 2}Cl{sub 2}, 0 deg. C, N{sub 2}, 4 h, 57%; b) NaN{sub 3}, reflux in H{sub 2}O/acetone for 22 h, 80%; c) PPh{sub 3}, THF/H{sub 2}O, 22 h, 79%; d) p-toluene sulfonyl chloride, Et{sub 3}N, 1,2-DCE, 22 h. Extraction: Sm(NO{sub 3}){sub 3}.6H{sub 2}O (2 mM) in 5 ml of aqueous NaOH (pH 10.5 - 14) + 6 ml of CH{sub 2}Cl{sub 2} solution of msa (20 mole equiv.) were rotated on a wheel (60 rpm; 20 h). Stripping: 5 ml of 0.1 M HNO{sub 3} + CH{sub 2}Cl{sub 2} solution of msa (after extraction, centrifugation and filtration) was rotated on the wheel (60 rpm; 20 h). Sm{sup 3+} was quantified using UV-Visible spectrophotometry.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

AGR-5/6/7 Partial Mass Balance

AGR-5/6/7 goals and test train configuration, status of AGR-5/6/7 mass balance work, preliminary results of Ag-110m, Cs-134, and Eu-154.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DoD Symposium Presentation

TRISO fuel, or TRi-structural ISOtropic, is a ceramic based nuclear fuel capable of operating at high temperatures (up to 1600°C). The fuel, consisting of uranium oxycarbide (UCO) fuel kernels, is coated with three layers of carbon and ceramic (dense silicon carbide) materials to capture and contain radioactive fission products. BWXT, located in Lynchburg, VA, has perfected the manufacturing techniques to produce this fuel and is currently the only US company licensed to produce this irradiation-tested fuel. BWXT is currently producing fuel in support of a demonstration reactor scheduled for startup on the INL site in 2025. The first fuel delivery, estimated at 200 kilograms High Assay Low Enriched Uranium, is targeted for the end of CY2024.

07 ISOTOPE AND RADIATION SOURCES↗