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At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Bayesian SegNet for Semantic Segmentation with Improved Interpretation of Microstructural Evolution During Irradiation of Materials

Understanding the relationship between the evolution of microstructures of irradiated LiAlO2pellets and tritium diffusion, retention and release could improve predictions of tritium performance. Given expert-labeled segmented images of irradiated and unirradiated pellets, we trained Deep Convolutional Neural Networks to segment images into defect, grain, and boundary classes. Qualitative microstructural information was calculated from these segmented images to facilitate the comparison of unirradiated and irradiated pellets. We tested modifications to improve the sensitivity of the model, including incorporating meta-data into the model and utilizing uncertainty quantification. The predicted segmentation was similar to the expert-labeled segmentation for most methods of microstructural qualification, including pixel proportion, defect area, and defect density. Overall, the high performance metrics for the best models for both irradiated and unirradiated images shows that utilizing neural network models is a viable alternative to expert-labeled images.

Oostrom, Marjolein T.↗
DOE OSTI2021

Alpha-decay induced shortening of fission tracks simulated by in situ ion irradiation

The diffusion of defects (e.g., vacancies and interstitials) and elements used for dating (e.g., He and Pb) in a mineral structure is a thermal process: This is the primary assumption used to determine the age and thermal history of minerals. For instance, thermal history reconstruction, through the number and length distribution of tracks produced by spontaneous fission of 238 U, is obtained by assuming a thermal event to be the only energy source for shortening of fission tracks. Here, we report a new, non-thermal energy source that induces additional shortening of fission tracks by the irradiation of alpha-recoils from the alpha-decay of 238 U and 232 Th. We simulate alpha-decay induced track-shortening by combining ion accelerator irradiations with transmission electron microscopy. This allows for the first observation of track-shrinkage during in situ ion irradiation. We show that rather than alpha-particles, alpha-recoils induce a significant shortening of fission tracks by nuclear-collisions. The shortening of track-length can be quantified as a function of alpha-decay event dose. However, apatite is less sensitive than zircon to this non-thermal process. Furthermore, the findings exemplify the interactions among different types of self-irradiation from alpha-particles, alpha-recoils and fission-fragment nuclei in single mineral grains and have important implications for the use of zircon and apatite for radiometric dating and thermochronology.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Cluster dynamics simulation of xenon diffusion during irradiation in UO 2

Diffusion of fission gas in UO 2 nuclear fuel impacts several important performance metrics, such as fission gas release, swelling, and thermal conductivity. Current empirical models of fission gas release have significant uncertainty, some of which derives from the bulk diffusion rate and its dependence on, for example, fuel chemistry and irradiation. In this work, we have applied the previously-developed Free Energy Cluster Dynamics (FECD) methodology in the code Centipede to calculate xenon cluster concentrations in UO 2 under intrinsic (high temperature) and irradiation-enhanced (intermediate temperature) conditions in order to develop a model of the xenon diffusion coefficient based on the atomic scale mechanisms responsible for transport. While the diffusion mechanism for xenon in UO 2 is adequately described by the Xe + U 2 O vacancy cluster for intrinsic conditions, a similar process is not capable of capturing measured in-pile fission gas diffusivity at intermediate temperatures. Therefore, a different diffusion mechanism must dominate under this regime. Using calculated atomistic data, we have shown that irradiation-enhanced diffusion at intermediate temperatures occurs via the larger Xe + U 4 O y vacancy clusters, which have lower migration barriers and increase in concentration by several orders of magnitude compared to intrinsic conditions. This mechanism is enabled by the increased uranium vacancy concentration under irradiation due to Frenkel pair production. In addition, the fast migration of uranium interstitials with two attached oxygen interstitials lowers the total uranium interstitial concentration through reactions with sinks. This allows the extended defects, such as Xe + U 4 O y vacancy clusters, to maintain high concentrations by limiting annihilation with attached vacancies. Furthermore, predictions using the Xe + U 4 O y diffusion mechanism are in good agreement with experiment, albeit with some differences in the Arrhenius slope, which we believe may be related to either experimental or model parameter uncertainty. Lastly, an analytical expression suitable for application in fuel performance simulations was derived to capture the predictions of the Centipede simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Electron tomography of unirradiated and irradiated nuclear graphite

Graphite is the moderator material of several Generation IV nuclear reactor concepts, as well as the British Advanced Gas-cooled Reactors (AGR). Porosity can heavily influence the material properties, mechanical irradiation response, and neutron induced shrinkage or swelling of nuclear-grade graphite. Due to the sub-micron size of several types of pores found in graphite, only a high-resolution imaging technique such as electron tomography are capable of visualizing these features in three dimensions. In this research, we used electron tomography to characterize as-received and neutron irradiated samples of IG-110 nuclear-grade graphite to show for the first time the 3D structure of both native and irradiation-induced nano-cracks. This technique also reveals unique characteristics of graphite such as the structure that surrounds pores and could be used to inform molecular dynamic simulations of irradiated graphite and experimental techniques such as gas-absorption. This work also shows the utility of this technique for the study of other nuclear porous carbon-based materials.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Characterization of mechanical properties and deformation behavior of highly irradiated 316L stainless steel from target modules at the Spallation Neutron Source using digital image correlation analysis

Tensile specimens with irradiation doses ranging from approximately 0.7 to 7.6 displacements per atom (dpa) were fabricated from 316L samples removed from Spallation Neutron Source (SNS) target vessels after service. Tensile tests were performed in conjunction with digital image correlation (DIC) characterization, which permitted detailed analysis of the strain distribution and deformation behavior of the specimen gauge sections during testing. The results show, while considerable hardening (500–700 MPa increase in yield strength) occurred in the 316L SNS target material with increasing dose, appreciable ductility (19–23% total elongation) remained in the material after irradiation to 7.6 dpa in the mixed neutron/proton environment of the SNS target region. The DIC data showed the strain distribution and deformation behavior of all irradiated specimens were nonuniform during testing. Strain localization in the form of deformation bands or “hotspots” was observed during testing, and in some specimens, the bands propagated as deformation waves along the gauge section. The “traveling wave” deformation behavior affected the amount of elongation each specimen achieved, and the scatter in the elongation data was attributed to the large variation in deformation behavior observed among different specimens. Bands of localized strain, or deformation bands, oriented at specific angles were observed in two-dimensional plots of the strain distribution in the gauge sections of some specimens. Overall, this paper describes the experimental procedures, results, and deformation behavior of tensile specimens from SNS targets after service and illustrates the value of DIC characterization during tensile testing of irradiated materials.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Thermal diffusivity and thermal conductivity of SiC composite tubes: the effects of microstructure and irradiation

Cladding thermal conductivity is an important physical property in assessing the performance of silicon carbide (SiC)-cladded fuels for nuclear reactors. However, there is a significant lack of reliable data, particularly for irradiated materials, because the geometry complicates the measurement. This study investigates the thermal diffusivity of coupons with a curvature, machined from SiC fiber–reinforced SiC matrix composite tubes, with and without neutron irradiation under light water reactor–relevant temperature and dose conditions. The tested materials included full composite and duplex SiC composite tubes. The measurements were conducted using a modern flash diffusivity apparatus. The analyzed area on the specimen during diffusivity testing was reduced for improved measurement accuracy due to sample curvature. Post-irradiation measurements showed that the effects of neutron irradiation on thermal conductivity (e.g. thermal defect resistivity) are different between SiC composite plates versus tubes. The difference was explained by higher matrix density of the tube than the plate. This study provides reliable thermal properties of prototypic SiC composite tubes useful for fuel performance modeling of SiC-based cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe 2+ up to 600 peak displacement-per-atom (dpa) at 450 °C. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. Furthermore, the comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Heavy ion irradiation effects on CrFeMnNi and AlCrFeMnNi high entropy alloys

Co-free but Al-included medium/high entropy alloys (M/HEAs) have gained increasing interests due to their lower cost and the potential to tune the multi-phase microstructure. The irradiation response of two Co-free HEAs, face-centered cubic (FCC) CrFeMnNi with limited Cr enriched α' phase and body-centered cubic (BCC) AlCrFeMnNi with B2s phase and nanoprecipitates were explored. Ion irradiations using 5 MeV Fe 2+ ions were performed at 500°C to a peak fluence of 50 and/or 100 displacements per atom (dpa). In dual-phase AlCrFeMnNi, there was no significant radiation induced segregation or chemical intermixing at the coherent matrix (FeCrMn-rich)/second phase (AlNi-rich) boundaries. In CrFeMnNi, limited voids were only detected at the peak damage location of ~ 50 dpa. On the other hand, voids were widely distributed in AlCrFeMnNi: under 50 and 100 dpa irradiation conditions, voids were found with larger dimension and denser distribution in the FeCrMn-rich matrix, smaller and slightly lower density in an AlNi-rich second phase. In addition, the diameter of the FeCMn-rich nanoprecipitates didn't reveal any tendency of dissolution or growth. This is correlated with their superior structural stability against irradiation. Significant radiation-induced hardening (increases from 3.8 ± 0.2 GPa to 4.7 ± 0.6 GPa) was measured in CrFeMnNi, but only ~ 4% hardness increase (from 7.4 ± 0.8 GPa to 7.7 ± 0.4 GPa) was noted in AlCrFeMnNi. Finally, in addition to the radiation-induced defects, such as voids, dislocation loops and point defects, other factors, such as chemical short-range ordering may play an important role.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Post-irradiation examination of low burnup U 3 Si 5 and UN-U 3 Si 5 composite fuels

Here this work presents post-irradiation examination data on UN-U 3 Si 5 and U 3 Si 5 fuels at low burnup (i.e., <10–15 GWd/tHM) with Kanthal AF® cladding. The results suggest good irradiation performance for both the silicide and nitride-silicide composite pellets. Optical microscopy revealed that the pellet-cladding gap is still open, and limited axial cracking was observed only in UN-U 3 Si 5 pellets. Microcracking was isolated to the U 3 Si 5 phase in all cases and was observed in pre-irradiation and depleted pellets, indicating that it was not irradiation induced. The fission gas release was minimal for the calculated fission density achieved (2.6 – 3.15 × 10 20 fiss/cm 3 ). No fission gas bubbles were observed in the optical metallography. These results suggest acceptable swelling and fission gas behavior for both the single phase and composite compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Contributions to the mechanistic understanding of the microstructural evolution in irradiated U-Mo dispersion fuel

Here, advanced microstructural characterization techniques, such as scanning electron microscopy (SEM) and scanning transmission electron microscopy - energy dispersive x-ray spectroscopy (STEM-EDS), were used to interpret the fuel microstructure evolution and fission products behavior in U-Mo dispersion fuel irradiated in the Advanced Test Reactor (ATR) as part of the European Mini-Plate Irradiation Experiment (EMPIrE) test. The larger as-fabricated fuel grain size achieved by heat-treating the U-Mo powder resulted in slower high burnup structure (HBS) development and reduced fission gas porosity. Slower HBS kinetics was observed at the fuel kernels’ periphery, which contained smaller and less fission gas bubbles at all fission densities (FDs) investigated and was attributed to a locally reduced damage density and fission products concentration, as corroborated with Monte Carlo simulations. The non-refined grains at the fuel kernel periphery hosted a perfectly ordered fission Gas Bubble Superlattice (GBS) up to 6.3 × 10 21 fissions/cm 3 . Nano-scale STEM-EDS analysis presented in this study provided useful information on the GBS characteristic morphology and evolution in U-Mo fuel. The concentration of fission gas in the GBS progressively increased with FD, pointing to an evolution of the nanobubble pressure status with irradiation. A possible connection between the GBS collapse and HBS onset is proposed for which there exists a threshold in the misorientation of the refined sub-grains above which the GBS stability during irradiation is no longer preserved, resulting in the GBS collapse.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten

The effect of mixed spectra and thermal neutron shielded irradiation on tungsten was evaluated with plasma exposure in the tritium plasma experiment followed by thermal desorption spectroscopy, X-ray photoelectron spectroscopy, and transmission electron microscopy. The two different irradiation campaigns were performed at the High Flux Isotope Reactor to 0.39–0.74 displacement per atom (dpa) in the 894–1379 K temperature range. A neutron spectrum influence on the void size and void number density was not observed. However, a strong correlation was found between void size and void number density with temperature, but not with dpa in the limited dpa range of this study. Thermal neutron shielding significantly reduced the transmutation to Re+Os. Higher irradiation temperature will lead to larger voids with lower number density, which reduces deuterium retention. In conclusion, grain growth was also observed for high-temperature irradiation of over ~1300 K within the limited grains visible in the transmission electron microscopy specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Understanding fission gas bubble distribution, lanthanide transportation, and thermal conductivity degradation in neutron-irradiated α-U using machine learning

U—10Zr based metallic nuclear fuel is the leading candidate for next-generation sodium-cooled fast reactors in the United States. US research reactors have used and tested this fuel type since the 1960s and accumulated considerable experience and knowledge about the fuel performance. Most of the knowledge, however, remains empirical. The lack of mechanistic understanding of fuel performance puts a large burden on proof through experimental verification for the qualification of U—10Zr fuel for commercial use. Further, this paper proposes an image data-driven machine learning approach, coupled with domain knowledge provided by advanced post irradiation examination, to provide unprecedented quantified insights into the morphology, size, density and the connectivity of fission gas bubbles and their effects on the fission product transportation and thermal conductivity. Specifically, we developed a method to automatically detect, extract statistics, and classify ~19,000 fission gas bubbles into different categories, and quantitatively link the data to lanthanide transportation through connected bubbles and degradation of thermal conductivity along the radial temperature gradient in a neutron irradiated U—10Zr annular fuel. Results indicate the approach can be modified to study other irradiation effects, such as secondary phase redistribution and gaseous fuel swelling in other irradiated nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Modeling of surface-state induced inter-electrode isolation of n -on- p devices in mixed-field and γ -irradiation environments

Position sensitive n-on-p silicon sensors will be utilized in the tracker and in the High Granularity Calorimeter (HGCAL) of the Compact Muon Solenoid (CMS) experiment at High Luminosity Large Hadron Collider (HL-LHC). The detrimental effect of the radiation-induced accumulation of positive net oxide charge on position resolution in n-on-p sensors has typically been countered by the application of isolation implants like p-stop or p-spray between n + -electrodes. In addition to the positively charged layer inside the oxide and close to the Si/SiO 2 -interface, surface damage introduced by ionizing radiation in SiO 2 -passivated silicon particle detectors includes the accumulation of trapped-oxide-charge and interface traps. A previous study of either n/γ (mixed field)- or γ-irradiated Metal-Oxide-Semiconductor (MOS) capacitors showed evidence of substantially higher introduction rates of acceptor- and donor-type deep interface traps (N it,acc/don ) in mixed-field environment. Here, in this work, an inter-pad and -strip resistance (or resistivity (ρ int )) simulation study of n-on-p sensors with and without p-stop isolation implants was conducted for both irradiation types. Higher levels of ρint showed correlation to higher densities of deep N it,acc/don , with the inter-pad isolation performance of the mixed-field irradiated sensors becoming independent of the presence of p-stop implant between the n + - electrodes up to about 100 kGy. The low introduction rates of deep N it,acc/don in γ-irradiated sensors resulted in high sensitivity of ρ int to the presence and peak doping of p-stop above the lowest dose of about 7 kGy in the study. As a consequence of the advantageous influence of radiation-induced accumulation of deep N it on the inter-electrode isolation, position sensitive n-on-p sensors without isolation implants may be considered for future HEP-experiments where the radiation is largely due to hadrons.

Akchurin, N. [Texas Tech Univ., Lubbock, TX (Unite↗
DOE OSTI2020

In-situ measurement of irradiation behavior in LiNbO 3

In-situ measurement of LiNbO 3 based surface acoustic wave (SAW) crystal resonator device under irradiation was demonstrated and used to characterize the impact of radiation on physical properties of this material. The resonant frequency of the SAW device was monitored as the output power of the reactor was varied. Upon step increase of the reactor power, a gradual shift in the device’s resonant frequency was observed. This frequency shift initially exhibits a linear growth and eventually reaches an equilibrium value proportional to the reactor power. The observed behavior can be attributed to two competing processes: increase of temperature due to gamma heating or accumulation of irradiation induced defects. In both cases, the response is attributed to changes in the physical properties of LiNbO 3 , particularly the elastic constants. This demonstrated ability to measure materials properties under irradiation is attractive for development of sensors and performing materials science under irradiation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Measurements of displacement cross sections of metals for 120-GeV proton beam irradiation

The number of displacements per atom (dpa) is widely used as an indicator of irradiation damage of materials in proton accelerator facilities. Experiments have been carried out to validate the dpa of metallic materials for protons with energies below 3 GeV. However, measurements of the displacement cross-sections for high-energy protons above 3 GeV have not been carried out and the calculations have not been validated. To validate the displacement cross section of metals in high-energy region, electrical resistivity changes in wires of aluminum, copper and tungsten at 8 K were measured using protons with energies of 120-GeV. The results show that the Norgett-Robinson-Torrens dpa model of the Particle and Heavy Ion Transport Calculation Code overestimates the experimental data. On the other hand, the calculated results using the athermal recombination corrected dpa model were in agreement with the measured displacement cross sections. In the proton energy region above 1 GeV, the displacement cross section is almost constant, which is due to the fact that the damage energy of the material under 1 GeV proton irradiation is almost the same as under 120 GeV proton irradiation. Damage recovery of defects accumulated in the sample was also measured using isochronal annealing: At 80 K, approximately 60 % and 80 % of the damage remains for copper and tungsten, respectively. These results are the same as those obtained from other experiments on proton and neutron irradiation.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Thermal hydraulic analysis of the AFIP-7 irradiation test in the Advanced test Reactor – Model correlation and performance evaluation

We report as a system level code, RELAP5 is widely used in the nuclear field for the reactor hydraulic analysis. In the component/experimental level, it is frequently employed as well. This paper demonstrates that the RELAP5 flow simulation of the AFIP-7 irradiation experiment which has a complicated 3D geometry/flow path deviates by ~28% from the flow evaluation through a computational fluid dynamics (CFD) simulation, which has been validated against a flow test performed at Oregon State University. Thermal safety compliance analysis of an experiment planned to be irradiated in the Advanced Test Reactor is usually performed using the finite element analysis code, Abaqus. This research reveals the disadvantages of the Abaqus simulation in the flow instability and departure from nucleate boiling evaluations via a conjugate heat transfer analysis of the AFIP-7 irradiation experiment. As a result, a detailed CFD simulation is suggested for irradiation experiments with complicated flow paths, rather than a simplified RELAP5 simulation or hand calculations. The validated simulation approach will be integrated with the Boehmite correlations to investigate the oxide growth prediction on the fuel cladding in Part II of this research.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Modeling and design of a separate effects irradiation test targeting fission gas release from Cr-doped UO 2

Fission gas release (FGR) from nuclear fuel during operation can diminish heat transfer properties across the pellet-cladding gap and increase the fuel rod internal pressure, thereby posing a concern to fuel reliability and safety during an accident. Enlarging the fuel grain size, which has been shown to improve fission gas retention, can be achieved by doping the fuel feedstock prior to sintering. In this work, the BISON fuel performance code was used to predict FGR from undoped and chromia-doped UO 2 (referred to as Cr-doped UO 2 ) fuel specimens with different grain sizes and across various temperatures. The BISON models identified the irradiation conditions for which FGR is most significant, and a separate effects irradiation experiment in the High Flux Isotope Reactor (HFIR) was then developed targeting those conditions. Further, the experiment leveraged the MiniFuel irradiation capability at Oak Ridge National Laboratory and consisted of 12 fuel specimens of varying grain size and Cr content. A coupling scheme between BISON FGR results and the ANSYS finite element thermal model used for experiment design was formulated to predict cumulative FGR from each fuel specimen based on expected irradiation temperature histories. The fuel samples were fabricated and characterized as a part of this work, and the fuel compositions modeled in BISON were representative of the specimens used in the experiment. This combined modeling and experimental effort aims to study the effect of fuel grain size and Cr content on FGR and to provide simulated BISON FGR results that can be used for future model validation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Beyond interpolation: Physics-inspired gating transformers for extrapolating irradiation conditions to novel nuclear fuels

The qualification of advanced nuclear fuels relies on irradiation experiments in test reactors that emulate commercial conditions. Designing these tests requires accurate prediction of key irradiation quantities, particularly heat generation rate and burnup, yet obtaining them typically involves computationally expensive multi-step simulation workflows. We propose a physics-inspired gating transformer (PIGT) that integrates an inverse-square, distance-based attenuation into the encoder representation to bias attention toward physically relevant spatial relationships while retaining data-driven flexibility. Using MiniFuel irradiation data from the High Flux Isotope Reactor at Oak Ridge National Laboratory, we benchmark against ensemble methods, feedforward and recurrent networks, convolutional models, and standard transformers. While baseline models perform well under interpolation, they exhibit a pronounced generalization gap when evaluated on fuels not included in the training set. The proposed model consistently improves extrapolative accuracy and stability, yielding the strongest performance on unseen fuel configurations. These results indicate that a lightweight physics structure embedded within attention mechanisms can substantially improve robustness, enabling more reliable surrogate predictions to accelerate the design of nuclear fuel irradiation experiments.

Fuel qualification↗