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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Impact of post-irradiation annealing on mechanical performance of irradiated 718 alloy

Here, the effect of post-irradiation annealing on solution-annealed 718 alloy was investigated using advanced mechanical testing, fractography, scanning electron microscopy, and transmission electron microscopy. Specimens were extracted from a proton beam window operated at the Spallation Neutron Source, irradiated with 940 MeV protons to a maximum dose of approximately 9.7 displacements per atom (dpa) at a calculated temperature not exceeding 110 °C while in service. Helium and hydrogen concentrations reached about 1700 and 6900 atomic parts per million (appm), respectively. Despite irradiation and high tensile strength (yield stress over 1 GPa), the material exhibited significant ductility. Annealing at 500 °C, 700 °C, and 900 °C for 30 min resulted in an appreciable decrease in yield strength and an increase in ductility for annealing treatments at 500 °C and 900 °C relative to the strength and ductility of the as-irradiated material. The presence of helium and hydrogen led to cavity formation and cleavage-like brittle features on fractured surfaces; however, high-magnification imaging revealed the presence of small-scale ductile dimples, indicating that the fracture mechanism remained mixed. The annealed specimens retained total elongation levels of 14–33 %, and there was no sudden drop in ductility after heat treatments. The ductility level in the irradiated and annealed material is notable despite the presence of helium and hydrogen.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Temperature dependence of irradiation-induced nanocrystallization in amorphous silicon carbide

Ion irradiation induced nucleation and growth in amorphous SiC at elevated temperatures are investigated in this study. Both as-deposited amorphous and Kr ion irradiation-amorphized SiC films are used for the investigation. Similar behavior of particle precipitation in the two types of the films is observed. It is found that the threshold temperature for nucleation is between 550 and 700 K, which is considerably higher than the critical temperature for full amorphization in SiC irradiated with Kr ions. There is a temperature regime up to at least 900 K, where the growth rates are only weakly dependent of the irradiation temperature. This temperature regime is potentially useful to tailor the size and density of nanocrystalline SiC precipitates by optimizing the ion irradiation conditions.

Ion irradiation, Nucleation and Growth, Silicon Ca↗
DOE OSTI2025

Accelerated Irradiation Testing and Post-Irradiation Characterization: U.S.-Based Capabilities for Advanced Nuclear Systems and Radioisotope Production

Irradiation experiments and post-irradiation examinations, together referred to as irradiation testing (IRT), are prerequisites for nuclear fuel and material qualification for the deployment of new and advanced reactors, as well as radioisotope production, thereby ensuring regulatory compliance. Qualified research and test reactors (RTRs) and testing facilities are essential to enable IRT to verify performance and safety under prototypical reactor conditions. In the past, qualification of new fuels or structural materials required about 20 years. Synergist strategies, advanced tools, and qualified methods are needed to greatly reduce this timeframe of IRT and radioisotope production. This study, termed accelerated-IRT, focuses on identifying gaps and leveraging U.S.-based RTRs and material testing capabilities, leveraging the preliminary evaluation and qualification of selected RTRs to provide a generic as well as specific-case solution paths forward, ensuring adherence to stringent regulatory standards. Furthermore, IRT and radioisotope production utilizing qualified RTRs necessarily includes modeling and simulation to support the design (i.e. neutronics, thermal, and structural aspects) and manufacturing of irradiation test specimens, vehicles, capsules, apparatuses, and flow loops. In addition, IRT can be improved by applying advanced manufacturing techniques and in-pile sensors and instrumentation, as discussed in this study.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Isotope-Specific Analysis of Neutron-Irradiated Lithium Aluminate Ceramics by Nuclear Magnetic Resonance Spectroscopy

The effects of neutron irradiation on lithium aluminate targets and the permeation of tritium into surrounding components are primary concerns in the development of materials for use in tritium production. The high specific activity and volatility of the tritium in activated samples greatly increase the challenges of experimental analysis. This paper reports the determination of the chemical forms and quantities of H-3 and Li-6 in irradiated and non-irradiated LiAlO2 by nuclear magnetic resonance (NMR) spectroscopy, and demonstrates that measurements can be performed in a non-destructive, safe manner to obtain isotope-selective, quantitative information on highly hazardous materials.

Tritium, solid state nuclear magnetic resonance sp↗
DOE OSTI2024

Estimating UV-B, UV-Erithemic, and UV-A Irradiances From Global Horizontal Irradiance and MERRA-2 Ozone Column Information

The ground ultraviolet (UV) solar radiation is relevant due to its impacts on plastics degradation (mainly UVA) and on human health (UVB and erithemic UV (UVE)). UV ground measurements are not as ubiquitous as the relatively common global horizontal irradiance (GHI) measurements. Three simple models that estimate the UVA, UVB, and UVE components of solar irradiance from GHI and ozone column information are locally adjusted and validated. Five one-minute datasets from three sites in southeastern South America and two in the United States are used for simultaneous solar irradiance and UV data. All sites correspond to temperate mid-latitude regions. Simultaneous atmospheric total ozone column information is obtained from the reanalysis modern-era retrospective analysis for research and applications (MERRA-2) database for each site. Aside from locally adjusted models, average models with a single set of coefficients are also evaluated. For instance, the best average model is able to estimate UVE with a typical uncertainty below 12% and mean biases between +-3%, relative to the average of the measurements. Similar results are reported for the UVB and UVA components. These results, which can be useful in regions with similar climate and geography, provide a simple way to estimate UV irradiance under all-sky conditions with known uncertainty. This is an alternative to global satellite-based UV estimates, which can have high uncertainties at specific locations. Because MERRA-2 information has a global coverage, when coupled with good satellite-based estimates for GHI, UV irradiances can be estimated by this method over a large territory.

environmental UV radiation↗
DOE OSTI2022

Irradiated stainless steel 347 and irradiated, welded, and hydrogen charged Zircaloy-4 property data for the target solution vessel and support lines of SHINE

In support of the reactor-vessel development effort of SHINE medical technologies, irradiation of its structural materials and post-irradiation testing have been performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This irradiation was performed on welded and unwelded AISI 347 steel and Zircaloy-4 samples at low temperature (~60°C) in contact with the reactor coolant. Hardness, Charpy, and tensile data have been taken from all cases of unirradiated specimens, while post-irradiation tensile and hardness data have been taken from AISI 347 base metal and Zircaloy-4 base metal and weld specimens so far. Additionally, 3-point bend data have been taken from explosion-welded specimens. The expected changes to hardness, yield strength, ultimate tensile strength, uniform elongation, and total elongation as a result of the low temperature and low dose neutron irradiation are reported here.

36 MATERIALS SCIENCE↗
DOE OSTI2023

AFC FY 2023 HFIR Irradiation Test Matrix – Supported by the Design of a Ring Specimen Irradiation Vehicle

The development and deployment of accident-tolerant cladding materials in light water nuclear reactors requires the collection of neutron-irradiated properties of these materials with specific consideration to the material properties orientation (i.e., tube geometry). To that end, a test plan to irradiate several cladding tube and ring specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was proposed, along with the development of a new capsule design that features ring specimens with a preexisting gauge section. This ring specimen geometry removes the need to machine a gauge section on the irradiated specimen and thus allows for direct application of a ring-pull test post-irradiation. The test matrix includes uncoated and chromium-coated Zircaloy-4 and advanced zirconium alloy cladding specimens, pre-characterized at ORNL. Plans are to irradiate the specimens at about 4 and 10 dpa—corresponding to two and five cycles at the HFIR centerline—with a target specimen gauge section temperature of 330°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Correlation of irradiation data using activation fluences and irradiation temperature.

Nil-ductility transition temperature data for A302-B ferritic steel were analyzed by using multiple regression. Four independent variables were used - i.e., specimen temperature during irradiation, and time-integrated specific activations (activation fluences) of fast-, intermediate-, and thermal-neutron detectors. The results of these analyses were predictive equations having the increase in transition temperature as the dependent variable. For a reference point, a similar analysis was also performed using the flux integral above 1 MeV and the irradiation temperature as independent variables. Also, the effect of excluding irradiation temperature from these analyses was studied. The results of the analyses showed that excellent correlation was achieved by using activation fluences and irradiation temperature as the independent variables. Irradiation temperature was found to account for from 29 to 46% of the data variability depending on the model used.

Lynch, J. H.↗
NASA NTRS1982

Irradiation effects in rapidly and conventionally solidified alloys. Phase stability in rapidly solidified N i-Nb under Ni ion irradiation

Two alloy compositions in the Ni-Nb system (Ni60Nb40 and Ni85Nb15) were produced by rapidly quenching from the melt with the piston anvil technique. The Ni60Nb40 was transformed to a metastable, partially crystalline state by heat treatment in a differential scanning calorimeter. The Ni85Nb15 was fully crystalline, with the majority of the grains composed of collections of primary dendrite arms. Both compositions were irradiated with 4 MeV Ni++ ions. The irradiation induced microstructures were examined by transmission electron microscopy and compared with thermally aged samples. The thermal evolution was arrested by ion irradiation in the temperature range studied, by inhibiting the nucleation of the NiNb phase. No irradiation induced voids were observed. It is found that the ion irradiation drives the microstructure along a different path than thermal evolution.

Source record↗
NASA NTRS2002

Response of thyroid follicular cells to gamma irradiation compared to proton irradiation: II. The role of connexin 32

The objective of this study was to determine whether connexin 32-type gap junctions contribute to the "contact effect" in follicular thyrocytes and whether the response is influenced by radiation quality. Our previous studies demonstrated that early-passage follicular cultures of Fischer rat thyroid cells express functional connexin 32 gap junctions, with later-passage cultures expressing a truncated nonfunctional form of the protein. This model allowed us to assess the role of connexin 32 in radiation responsiveness without relying solely on chemical manipulation of gap junctions. The survival curves generated after gamma irradiation revealed that early-passage follicular cultures had significantly lower values of alpha (0.04 Gy(-1)) than later-passage cultures (0.11 Gy(-1)) (P < 0.0001, n = 12). As an additional way to determine whether connexin 32 was contributing to the difference in survival, cultures were treated with heptanol, resulting in higher alpha values, with early-passage cultures (0.10 Gy(-1)) nearly equivalent to untreated late-passage cultures (0.11 Gy(-1)) (P > 0.1, n = 9). This strongly suggests that the presence of functional connexin 32-type gap junctions was contributing to radiation resistance in gamma-irradiated thyroid follicles. Survival curves from proton-irradiated cultures had alpha values that were not significantly different whether cells expressed functional connexin 32 (0.10 Gy(-1)), did not express connexin 32 (0.09 Gy(-1)), or were down-regulated (early-passage plus heptanol, 0.09 Gy(-1); late-passage plus heptanol, 0.12 Gy(-1)) (P > 0.1, n = 19). Thus, for proton irradiation, the presence of connexin 32-type gap junctional channels did not influence their radiosensitivity. Collectively, the data support the following conclusions. (1) The lower alpha values from the gamma-ray survival curves of the early-passage cultures suggest greater repair efficiency and/or enhanced resistance to radiation-induced damage, coincident with the expression of connexin 32-type gap junctions. (2) The increased sensitivity of FRTL-5 cells to proton irradiation was independent of their ability to communicate through connexin 32 gap junctions. (3) The fact that the beta components of the survival curves from both gamma rays and proton beams were similar (average 0.022 +/- 0.008 Gy(-2), P > 0.1, n = 39) suggests that at higher doses the loss of viability occurs at a relatively constant rate and is independent of radiation quality and the presence of functional gap junctions.

Non-NASA Center↗
DOE OSTI2025

Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance

Here, the development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.

BISON↗
DOE OSTI2026

Irradiation-induced grain boundary migration in an ion irradiated Fe-Ni-Cr-Mn alloy

Irradiation-induced grain boundary migration (RIGM) has previously been observed in different metallic systems. However, the mechanisms and dependence on irradiation conditions remained unclear. Here, we report on radiation-induced GB migration (RIGM) in a coarse-grain Fe-Ni-Cr-Mn austenitic alloy irradiated with Fe ions at 400 °C to 15, 35, and 70 dpa. Grain boundary migration is only observed in the irradiated regions and exhibits a linear dependence on the locally delivered dose, as estimated from SRIM calculations. Notably, the same linear dependence is noted across the nominal doses. Furthermore, the evidence points to grain boundary migration being mediated by the flux of point defects to grain boundaries and motivates additional systematic studies of RIGM in these systems.

Grain boundary↗
DOE OSTI2021

Helium bubble nucleation and growth in alloy HT9 through the use of in situ TEM: Sequential he-implantation and heavy-ion irradiation versus dual-beam irradiation

The formation of He bubbles in Ferritic/Martensitic steel HT9 is investigated through the use of in situ Transmission Electron Microscopy coupled with He implantation and heavy ion irradiation. Of particular interest is the effect of increasing He appm/dpa ratio on the formation and growth of the bubbles, as well as the effect of the sequential order of ion irradiation i.e. He-pre-implantation followed by heavy-ion irradiation versus true dual-beam irradiation. Finally, the role of He is discussed.

Duemmler, Kai↗
DOE OSTI2020

A thermodynamics-based damage model for the non-linear mechanical behavior of SiC/SiC ceramic matrix composites in irradiation and thermal environments

A damage model is developed and validated with experimental data for the non-linear mechanical behavior of SiC/SiC composite materials in nuclear applications. Cyclic thermal and mechanical loading associated with neutron irradiation effects of these composites leads to wide-spread and progressive micro-cracking that leads to loss of thermal conductivity and further enhancement of thermo-mechanical damage. A physics-based model of wide-spread micro-cracking is developed within the thermodynamic framework of continuum damage mechanics. Evolution equations for damage parameters that describe the growth of continuum damage are developed, where the material variables are obtained from experiments. The model novelty is in coupling mechanical, thermal, and irradiation damage through a consistent thermodynamic framework, including loss of thermal conductivity due to the evolution of mechanically induced micro-cracks. A number of thermo-mechanical experiments were conducted to confirm model assumptions. The model is shown to be validated with out-of-pile experiments, and then implemented using commercial finite element code COMSOL to the fuel cladding problem with normal and high radiation dose cases.

Materials Science↗
DOE OSTI2025

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO2), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO 2 ), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1982

Active-region evolution and solar rotation variations in solar UV irradiance, total solar irradiance, and soft X rays

Variations in the total solar irradiance, solar UV spectral irradiance, and solar soft X-ray emission caused by active region evolution and solar rotation are analyzed by using concurrent measurements from the NIMBUS 7 and GOES satellites. The observations are interpreted by using simple empirical models that relate ground-based observations of the size and location of sunspots and plages to the full-disk temporal variations. It is found that the major dips in the photospheric total solar irradiance S, which are evident in both satellite measurements and model predictions, are usually not accompanied by outstanding enhancements in the chromospheric and upper photospheric UV spectral irradiance or coronal X rays. The main cause of this difference between the variability of S and of the UV flux is that the total chromospheric plage enhancements are not outstanding at those times when the total sunspot are outstanding. X rays are even more variable because of a much wider CMD sensitivity.

Donnelly, R. F.↗