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At least 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

WE-EF-BRA-10: Prophylactic Cranial Irradiation Reduces the Incidence of Brain Metastasis in a Mouse Model of Metastatic Breast Cancerr

Purpose: Prophylactic cranial irradiation (PCI) is a clinical technique used to reduce the incidence of brain metastasis and improve overall survival in select patients with acute lymphoblastic leukemia and small-cell lung cancer. We examined whether PCI could benefit breast cancer patients at high risk of developing brain metastases. Methods: We utilized our mouse model in which 500k green fluorescent protein (GFP)-labeled breast cancer cells injected into the tail vein of SCID/Beige mice resulted in brain metastases in approximately two-thirds of untreated mice. To test the efficacy of PCI, one set of mice was irradiated five days after cell injection with a single fraction of 4-Gy (two 2-Gy opposing fields) whole-brain irradiation on the XRAD 225Cx small-animal irradiator. Four controls were included: a non-irradiated group, a group irradiated two days prior to cell injection, and two groups irradiated 3 or 6 weeks after cell injection. Mice were sacrificed four and eight weeks post-injection and were evaluated for the presence of brain metastases on a fluorescent stereomicroscope. Results: The incidence of brain metastasis in the non-irradiated group was 77% and 90% at four and eight weeks, respectively. The PCI group had a significantly lower incidence, 20% and 30%, whereas the other three control groups had incidence rates similar to the non-treated control (70% to 100%). Further, the number of metastases and the metastatic burden were also significantly lower in the PCI group compared to all other groups. Conclusion: The timing of irradiation to treat subclinical disease is critical, as a small dose of whole-brain irradiation given five days after cell injection abrogated tumor burden by greater than 90%, but had no effect when administered twenty-one days after cell injection. PCI is likely to benefit breast cancer patients at high risk of developing brain metastases and should be strongly considered in the clinic.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2026

HDG-1 Experiment Irradiation Monitoring Data Qualification Final Report

SUMMARY The U.S. Department of Energy (DOE) Advanced Reactor Technologies (ART) Graphite Research and Development (GRD) Program is conducting a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. This report documents the qualification of irradiation monitoring data for the fifth experiment, High Dose Graphite-1 (HDG-1). Qualified monitoring data are required by the ART program to support the design and licensing of the first high-temperature reactor (HTR) nuclear plant. Data are classified as Qualified if they meet the usage requirements described in the experiment planning and quality assurance (QA) documents, Failed if they do not meet those requirements and provide no usable information, or Trend if they do not fully meet all requirements but still provide useful information subject to an assessment of how any deficiencies may affect a particular use of the data. HDG-1 irradiation began with Advanced Test Reactor (ATR) Cycle 168B on August 24, 2020, and concluded after Cycle 173C on January 27, 2025. The HDG-1 capsule was removed from the reactor core twice—during core internal change (CIC) Cycle 170A and powered axial locator mechanism (PALM) Cycle 172A—to prevent overheating of the graphite specimens during high-power PALM cycles. The capsule was therefore irradiated during a total of seven normal ATR cycles: 168B, 169A, 171A, 171B, 173A, 173B, and 173C. Irradiation monitoring data evaluated in this report include thermocouple (TC) temperature, gas flow rate, gas moisture, gas pressure, specimen load, and graphite stack displacement. Temperature. A total of 14,508,065 TC temperature records were captured. Of these, 13,901,785 (95.8%) are Qualified and 606,280 (4.2%) are Failed. The principal source of failed temperature data was the instrument failure of TC-9 (Zone 2) on June 24, 2024, and TC-10 (Zone 1) on July 5, 2024, near the end of Cycle 173A, which resulted in 595,554 Failed readings. An additional 379 missing values and 10,347 slightly negative values from TC-13 during ATR outages are also Failed. Neither TC-9 nor TC-10 was used as a temperature-control TC, and their failures did not compromise capsule condition monitoring. Correlation analysis of all 13 TCs found no evidence of virtual junction formation. Control chart analysis revealed clear downward drift of approximately 80°C for TC-6 (Zone 3) relative to other stable TCs, and possible downward drift of approximately 60°C for TC-13 relative to the Zone 5 control TC (TC-1), though TC-13 remained consistent with the Zone 2 control TC (TC-12). Gas flow. A total of 20,088,090 gas flow rate records were captured. Of these, 19,941,463 (99.3%) are Qualified and 146,627 (0.7%) are Failed due to missing values. All argon, helium, and total gas flow data were within expected ranges throughout the irradiation. Gas moisture. A total of 1,116,005 outlet gas moisture values were captured. Of these, 1,101,421 (98.7%) are Qualified and 14,584 (1.3%) are Failed, comprising 14,556 out-of-range values and 28 missing values. The out-of-range moisture values exceeded 22,000 ppmv for approximately 1 week at the beginning of Cycle 173A, when accumulated moisture evaporated after the capsule was retrieved from water storage during PALM Cycle 172A and reinserted into the east flux trap. Moisture levels returned to below 25 ppmv for the remaining three cycles, and the transient high-moisture event did not affect the integrity of specimen irradiation. Gas pressure. A total of 7,812,035 gas pressure values were captured. Of these, 6,642,048 (85.0%) are Qualified and 1,169,987 (15.0%) outlet pressure values are Failed, comprising 718,537 zero outlet pressure values due to sensor failure from Cycle 168B through Cycle 171B, 54,550 missing values, and 396,900 too-low outlet pressure values, ranging from 1.1 to 1.6 psia after sensor replacement during Cycle 173A. Load. A total of 6,696,030 load values were captured. Of these, 6,694,580 (99.98%) are Qualified and 1,450 (0.02%) are Failed due to missing values. Applied loads to the six specimen stacks were stable throughout the irradiation. Stack displacement. A total of 6,696,030 displacement values were captured. Of these, 5,713,297 (85.32%) are Qualified and 3,781 (0.06%) are Failed due to missing values. Stack displacement increased consistently throughout the irradiation, reaching approximately 3.08 in. for Channels 5 and 6 by the end of irradiation. 978,952 (14.62%) substantially elevated displacements observed for Channel 6 beginning in Cycle 171A and for Channel 5 beginning in Cycle 173A are assigned Trend status. Raising pressure. A total of 1,115,999 raising pressure values were captured. Of these, 1,115,430 (99.95%) are Qualified and 569 (0.05%) are Failed due to missing values. Ram pressure. A total of 6,696,030 ram pressure values were captured. Of these, 6,692,249 (99.95%) are Qualified and 3,484 (0.05%) are Failed due to missing values. Stack raising was perf

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

In-situ irradiation-induced studies of grain growth kinetics of nanocrystalline UO 2

The thermal conductivity of UO 2 fuel needs to be high enough to dissipate the heat generated from the fission reaction. Since grain size affects thermal conductivity and grain size can evolve with irradiation, it is critical to understand in-reactor UO 2 grain growth. Most studies of grain growth in UO 2 are based on thermally driven processes at elevated temperatures. However, studies have shown that grain growth can occur even at cryogenic temperatures by ballistic processes. Such irradiation-induced grain growth in UO 2 is yet to be studied. Advanced in-situ Kr ion irradiation and transmission electron microscopy were systematically performed on nanocrystalline UO 2 thin films at temperatures ranging from 50 K to 1073 K; grain growth was observed at all temperatures. A combination of manual and machine learning techniques was used to measure and plot grain size evolution against irradiation fluence at various irradiation temperatures. The machine learning method has significantly improved the analysis efficiency and reduced human labors. The grain diameter data were fitted using classical grain growth and thermal spike models to describe grain growth kinetics with and without irradiation effect. Grain growth during low temperature irradiation (≤ 475 K) can be well described by the thermal spike model. Above 475 K, there were additional thermally assisted processes that further accelerate the grain growth. At the highest irradiation temperature about 1075 K, both irradiation-induced dislocation loops and cavities/bubbles were observed to form in the UO 2 . In this report, the effects of irradiation-induced defects on grain growth kinetics are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A secretome analysis reveals that PPARα is upregulated by fractionated-dose γ-irradiation in three-dimensional keratinocyte cultures

Studies have shown that γ-irradiation induces various biological responses, including oxidative stress and apoptosis, as well as cellular repair and immune system responses. However, most such studies have been performed using traditional two-dimensional cell culture systems, which are limited in their ability to faithfully represent in vivo conditions. A three-dimensional (3D) environment composed of properly interconnected and differentiated cells that allow communication and cooperation among cells via secreted molecules would be expected to more accurately reflect cellular responses. Here, we investigated γ-irradiation–induced changes in the secretome of 3D-cultured keratinocytes. An analysis of keratinocyte secretome profiles following fractionated-dose γ-irradiation revealed changes in genes involved in cell adhesion, angiogenesis, and the immune system. Notably, peroxisome proliferator-activated receptor-α (PPARα) was upregulated in response to fractionated-dose γ-irradiation. This upregulation was associated with an increase in the transcription of known PPARα target genes in secretome, including angiopoietin-like protein 4, dermokine and kallikrein-related peptide 12, which were differentially regulated by fractionated-dose γ-irradiation. Collectively, our data imply a mechanism linking γ-irradiation and secretome changes, and suggest that these changes could play a significant role in the coordinated cellular responses to harmful ionizing radiation, such as those associated with radiation therapy. This extension of our understanding of γ-irradiation-induced secretome changes has the potential to improve radiation therapy strategies. - Highlights: • γ-irradiation induced changes of cell adhesion, angiogenesis, and immune system in secretome of 3D-cultured keratinocytes. • Peroxisome proliferator-activated receptor-α (PPARα) was upregulated in response to fractionated-dose γ-irradiation. • The known PPARα target genes were differentially regulated by fractionated-dose γ-irradiation.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

Thalidomide Ameliorates Inflammation and Vascular Injury but Aggravates Tubular Damage in the Irradiated Mouse Kidney

Purpose: The late side effects of kidney irradiation include vascular damage and fibrosis, which are promoted by an irradiation-induced inflammatory response. We therefore treated kidney-irradiated mice with the anti-inflammatory and angiogenesis-modulating drug thalidomide in an attempt to prevent the development of late normal tissue damage and radiation nephropathy in the mouse kidney. Methods and Materials: Kidneys of C57Bl/6 mice were irradiated with a single dose of 14 Gy. Starting from week 16 after irradiation, the mice were fed with thalidomide-containing chow (100 mg/kg body weight/day). Gene expression and kidney histology were analyzed at 40 weeks and blood samples at 10, 20, 30, and 40 weeks after irradiation. Results: Thalidomide improved the vascular structure and vessel perfusion after irradiation, associated with a normalization of pericyte coverage. The drug also reduced infiltration of inflammatory cells but could not suppress the development of fibrosis. Irradiation-induced changes in hematocrit and blood urea nitrogen levels were not rescued by thalidomide. Moreover, thalidomide worsened tubular damage after irradiation and also negatively affected basal tubular function. Conclusions: Thalidomide improved the inflammatory and vascular side effects of kidney irradiation but could not reverse tubular toxicity, which probably prevented preservation of kidney function.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2022

Electronic energy loss and ion velocity correlation effects in track production in swift-ion-irradiated LiNbO3: A quantitative assessment between structural damage morphology and energy deposition

The primary motivation for studying how irradiation modifies the structures and properties of solid materials involves the understanding of undesirable phenomena, including irradiation-induced degradation of components in nuclear reactors and space exploration, and beneficial applications, including material performance tailoring through ion beam modification and defect engineering. In this work, the formation mechanism of latent tracks with different damage morphologies in LiNbO 3 crystals under 0.09–6.17 MeV/u ion irradiation with an electronic energy loss from 2.6–13.2 keV/nm is analyzed by experimental characterizations and numerical calculations. Irradiation-induced damage is preliminarily evaluated via the prism coupling technique to analyze the correlation between the dark-mode spectra and energy loss profiles of irradiated regions. Under the irradiation conditions of different ion velocities and electronic energy losses, different damage morphologies, from individual spherical defects to discontinuous and continuous tracks, are experimentally characterized. During ion penetration process, the ion velocity determines the spatiotemporal distribution of deposited irradiation energy induced by electronic energy loss, meaning that the two essential factors including electronic energy loss and ion velocity co-affect the track damage. The inelastic thermal spike model is used to numerically calculate the spatiotemporal evolutions of energy deposition and the corresponding atomic temperature under different irradiation conditions, and a quantitative relationship is proposed by comparison with corresponding experimentally observed track damage morphologies. Additionally, the obtained quantitative relationship between irradiation conditions and track damage provides deep insight and guidance for understanding the damage behavior of crystal materials in extreme radiation environments and selecting irradiation parameters, including ion species and energies, for ion beam technique application in atomic-level defect manipulation, material modification, and micro/nanofabrication.

36 MATERIALS SCIENCE↗
DOE OSTI2026

ROADRUNNER uranium nitride MiniFuel: Experimental design, fabrication and pre-irradiation baseline characterization for accelerated burnup testing

Uranium nitride (UN) is a promising fuel candidate for advanced reactor systems owing to its high uranium density and thermal conductivity; however, its qualification remains constrained by the scarcity of well-controlled irradiation performance data. Here, to address this limitation, the ROADRUNNER (Research On ADvancing the peRformance of UraNium Nitrides in Extreme enviRonments) campaign employs the MiniFuel platform in the High Flux Isotope Reactor (HFIR) to enable accelerated burnup irradiation testing under tightly controlled and largely isothermal conditions. This paper presents the experimental design, fuel fabrication, and pre-irradiation baseline characterization of the ROADRUNNER UN MiniFuel campaign. Thirty-six UN minidisc specimens were fabricated with systematically varied as-fabricated density (86–96% of theoretical density), carbon impurity content (961–5240 ppm), oxygen content (≤ ∼2000 ppm), and grain size (2.5–24 μm). The irradiation matrix spans nominal fuel temperatures of 873 K, 1173 K, and 1473 K and target burnups of 3.75%, 6.0%, and 7.5% fissions per initial metal atom (FIMA). Neutronic and thermal analyses were performed to define specimen-specific burnup accumulation and temperature histories, establishing the boundary conditions for subsequent in-pile behavior. Comprehensive pre-irradiation characterization—including dimensional metrology, density verification, impurity analysis, X-ray diffraction, Raman spectroscopy, scanning electron microscopy, X-ray computed tomography, and confocal profilometry—provides a detailed baseline for post-irradiation examination. Pre-irradiation data were further used to generate predictive estimates of fission gas release and swelling using existing empirical correlations. This quantitative comparison reveals substantial inter-model divergence at intermediate and elevated temperatures that exceeds propagated input uncertainties, highlighting structural gaps in the historical irradiation database. The ROADRUNNER irradiation campaign is currently underway in HFIR, with initial firs cycle completed in late 2025 and remaining targets scheduled through 2027. The experimental design and baseline dataset presented here establish the framework needed to interpret forthcoming post-irradiation measurements and to provide discriminating data for the validation and refinement of physics-based UN fuel performance models.

Lopes, Denise Adorno [Oak Ridge National Laborator↗
DOE OSTI2025

Tensile properties of the neutron irradiated HT-9 ACO-3 duct

HT-9, a ferritic-Martensitic steel, has attracted significant interest as a structural and cladding material for a variety of high dose nuclear component applications. This study examines the tensile properties of a HT-9 fuel assembly duct from the fast flux test facility, FFTF. The duct, from the Advanced Core Oxide-3 (ACO-3) irradiation test, was subjected to fast neutron irradiation exposure ranging from 3 dpa above the core top to nearly 155 dpa near the core mid-plane, spanning a temperature range from 382 °C at the lower end to 504 °C at the top. Due to the combination of thermal and neutron flux gradients along the core height. This yields a variety of irradiation damage processes which depend on core location. Tensile properties of specimens extracted at different positions along the height of the irradiated duct were measured at 25 °C, 200 °C and near their in-core irradiation temperature. The results show temperature has a greater impact than dose on post-irradiation strength and ductility. Two irradiation temperature regimes were determined, below irradiation temperatures of 420 °C irradiation hardening increases as irradiation temperature decreases, and above that temperature relatively minor changes are observed in tensile properties. These results are discussed in terms of expected microstructural changes, indicating that dislocation content is a strong contributor in these findings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

SU-E-T-296: Dosimetric Analysis of Small Animal Image-Guided Irradiator Using High Resolution Optical CT Imaging of 3D Dosimeters

Purpose: To verify the dosimetric characteristics of a small animal image-guided irradiator using a high-resolution of optical CT imaging of 3D dosimeters. Methods: PRESAEGE 3D dosimeters were used to determine dosimetric characteristics of a small animal image-guided irradiator and compared with EBT2 films. Cylindrical PRESAGE dosimeters with 7cm height and 6cm diameter were placed along the central axis of the beam. The films were positioned between 6×6cm{sup 2} cubed plastic water phantoms perpendicular to the beam direction with multiple depths. PRESAGE dosimeters and EBT2 films were then irradiated with the irradiator beams at 220kVp and 13mA. Each of irradiated PRESAGE dosimeters named PA1, PA2, PB1, and PB2, was independently scanned using a high-resolution single laser beam optical CT scanner. The transverse images were reconstructed with a 0.1mm high-resolution pixel. A commercial Epson Expression 10000XL flatbed scanner was used for readout of irradiated EBT2 films at a 0.4mm pixel resolution. PDD curves and beam profiles were measured for the irradiated PRESAGE dosimeters and EBT2 films. Results: The PDD agreements between the irradiated PRESAGE dosimeter PA1, PA2, PB1, PB2 and the EB2 films were 1.7, 2.3, 1.9, and 1.9% for the multiple depths at 1, 5, 10, 15, 20, 30, 40 and 50mm, respectively. The FWHM measurements for each PRESAEGE dosimeter and film agreed with 0.5, 1.1, 0.4, and 1.7%, respectively, at 30mm depth. Both PDD and FWHM measurements for the PRESAGE dosimeters and the films agreed overall within 2%. The 20%–80% penumbral widths of each PRESAGE dosimeter and the film at a given depth were respectively found to be 0.97, 0.91, 0.79, 0.88, and 0.37mm. Conclusion: Dosimetric characteristics of a small animal image-guided irradiator have been demonstrated with the measurements of PRESAGE dosimeter and EB2 film. With the high resolution and accuracy obtained from this 3D dosimetry system, precise targeting small animal irradiation can be achieved.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2019

NEUTRON-IRRADIATION–INDUCED DAMAGE ASSESSMENT IN CONCRETE USING COMBINED PHASE CHARACTERIZATION AND NONLINEAR FAST FOURIER TRANSFORM SIMULATION

Understanding irradiation effects in concrete cavities of light water reactors is critical to the long-term operation (60+ years) of US nuclear power plants. Susceptibility of concrete to neutron irradiation greatly varies as a function of its constituents: coarse aggregates, sand, and hardened cement paste (hcp). In particular, higher irradiation–susceptibility of aggregates was found to be a direct function of radiation-induced volumetric expansion (RIVE), or the propensity of aggregates to swell as a function of mineral contents, structures and textures. Irradiation-induced amorphization is accompanied by significant swelling, especially in silicates. For example, the maximum volumetric expansions of quartz and feldspars have been shown to be as large as ≈ 18% and ≈ 8%, respectively, while the change of density in calcite remains rather low (≈ 0.3–0.5%). High swelling is associated with a high number of covalent bonds (Si–O) and a high degree of polymerization of [SiO4] tetrahedrons. Differential RIVEs between rock-forming minerals cause high stresses and may potentially develop cracking within the aggregates at the interface between aggregates and the hcp and within the hcp. Modeling neutron-irradiation effects in concrete requires addressing the problem at a scale that captures both the details of the aggregates’ mineralogy and texture, which is <~ 100 μm, and the representative size of concrete, which is ~ 10 cm. This requires high-resolution, large mapping capabilities. To this end, the US Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program developed an original tool suite called Microstructure Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) in which Nonlinear 2D-fast Fourier transform (FFT) techniques are employed to model mechanical or irradiation-induced cracking of high-resolution complex microstructures derived from combined mapping techniques, including micro x-ray fluorescence (m-XRF), energy-dispersive x-ray spectroscopy (EDS), electron backscatter diffraction (EBSD), ellipsometry, and scanning electron microscopy (SEM). Irradiated mineral properties are derived from the Irradiated Minerals, Aggregate and Concrete (IMAC) database, which includes available relevant literature data. Rigorous coupled damage-creep simulations are obtained through sequential linear analysis. Illustrations of actual accelerated irradiation-damage propagation in a complex microstructures are presented herein.

Le Pape, Yann↗
DOE OSTI2016

Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and grain boundaries in the ion-irradiated alloys. More significant segregation was observed in the neutron irradiated alloys. For the more concentrated alloys, irradiation did not affect existing Cr segregation to grain boundaries and segregation to dislocation loops was not observed perhaps due to a change in the dislocation loop structure with increasing Cr concentration. Precipitation of α’ was observed in the neutron irradiated alloys containing over 9 at.% Cr. However ion irradiation appears to suppress the precipitation process. Initial low dose ion irradiation experiments strongly suggest a cascade recoil effect. The systematic analysis of grain boundary orientation on Cr segregation was significantly challenged by carbon contamination during ion irradiation or by existing carbon and therefore carbide formation at grain boundaries (sensitization). The combination of the proposed systematic experimental approach with atomistic modeling of diffusion processes directly addresses the programmatic need for developing and benchmarking predictive models for material degradation taking into account atomistic kinetics parameters

36 MATERIALS SCIENCE↗
DOE OSTI2017

AGC-3 Specimen Post-Irradiation Examination Data Package Report

The Advanced Reactor Technologies Graphite research and development program is conducting an extensive graphite irradiation program to provide data to assist in licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.[ ] Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be a critical parameter in determining the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. A second, much more significant, internal stress is experienced within the graphite components due to irradiation induced dimensional changes. These induced stresses result from neutron irradiation causing physical changes in the graphite structure, first shrinking it and then expanding it, with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear graphite grades is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Phase Transformation of Thermally Aged and Neutron Irradiated Duplex Stainless Steel Used in LWRs. Final report

The lifetime of reactor components made of duplex stainless steels can be limited by the embrittlement from thermal aging, neutron irradiation or a synergistic effect between thermal aging and neutron irradiation. There is still a large scientific knowledge gap in understanding the phase evolution in duplex stainless steels upon thermal aging with or without neutron irradiation, the synergistic effect between thermal aging and neutron irradiation, and the impacts of structural evolution on material mechanical responses. In this project, totally, five major tasks were completed through this project: a) the in-situ WAXS tensile tests of thermally aged cast duplex stainless steels, b) the EXAFS investigation of thermally aged cast austenitic stainless steel and welds of austenite sainless steel, c) the characterization of irradiation effect on neutron irradiated CASS using APT, d) characterize the ferrite hardening of neutron irradiated CASS using nanoindentation, and e) the FEM modelling to validate the in-situ WAXS tensile tests and to simulate the stress/strain distribution during deformation for thermally aged duplex stainless steel. In summary, the thermal aging without neutron irradiation can significantly increase the strength of ferrite phase through the spinodal decomposition and G-phase precipitations. The in-situ X-ray experiments show that the hardening effect in ferrite phase can significantly alter the load partition between the ferrite and austenite phases. Atom probe tomography characterizations on samples irradiated to 0.08, 3, 5, 10, 20, and 40 dpa shows that neutron irradiation has dominated the microstructural evolution after 3 dpa, and the spinodal decomposition has saturated yet even at 40 dpa based on the quantified wavelength and amplitude.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Summary of the Postirradiation Examination of the First Samples from a MiniFuel Irradiation

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly screen different nuclear fuel concepts. Observations on fission gas release, irradiation conditions and microstructure of the irradiated fuel show good fuel performance at the low burnup achieved in this initial irradiation. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU as was achieved after only 68 effective full power days of irradiation. Results from silicon carbide thermometry measurements further benchmarked the MiniFuel irradiation vehicle and indicated fuel was irradiated at temperatures ranging from 410-460°C. Extensive microstructural characterization on the irradiated fuel was performed and no significant irradiation induced changes were observed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Impact of Irradiation and Decay Time on Chemical Speciation in Uranyl Sulfate Solution

A pH-1, 200-g U/L, highly enriched uranyl sulfate solution was neutron-irradiated using an accelerator-driven neutron source (the Universal Neutron Irradiator, UNI) to fission U-235 under conditions similar to those used by SHINE to produce Mo-99 in a subcritical assembly. We completed a series of five uranyl sulfate irradiations to assess the impact of irradiation and decay time on uranium precipitation and iodine speciation in solution. The duration of each irradiation was increased in order, with the shortest irradiation taking 2.9 hours and the longest, 64.5 hours. A total of 9.5E16 U-235 nuclei were fissioned in all irradiations, and a maximum fission power density of 0.26 W/mL was reached in the final irradiation. We observed no evidence of uranyl peroxide or other precipitates, demonstrating that the addition of Fe 2+ to a final Fe concentration of 200 ppm in the target solution was successful at preventing precipitation under representative irradiation conditions. Nonvolatile I - was found to be the dominant (>70%) iodine species in solution, both immediately after irradiation and after days to weeks of decay time.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Mechanical Properties of Additively Manufactured 316L Stainless Steel Before and After Neutron Irradiation–FY23

This report presents the observed mechanical behavior of the additively manufactured (AM) 316L stainless steel (SS) before and after neutron irradiation. In the Advanced Materials and Manufacturing Technologies (AMMT) program, a variety of mechanical and physical property data are generated and accumulated to assess the AM austenitic alloy for nuclear reactor applications. The testing and evaluation task in the FY 2023 focused on elucidating the effects of sampling location and build size on the mechanical properties of AM 316L SS (in stress-relieved condition) before and after neutron irradiation. The laser powder bed fusion (LPBF) process produced 316L plates of three distinct sizes from which SS-J3 miniature tensile specimens were machined from six different locations. The tensile specimens were irradiated in the High Flux Isotope Reactor (HFIR) normally to 2 and 10 dpa at the target temperatures of 300 °C and 600 °C. Post-irradiation tensile testing was performed at room temperature, 300 °C, and 600 °C. The mechanical properties of AM 316L SS were significantly influenced by the characteristic microstructures of printed materials, which include fine grains and high-density dislocations. Compared with the traditional 316L SS, AM 316L showed higher initial strength and lower ductility. Regardless of sampling location, the AM 316L steel retained relatively high strength and ductility to the highest irradiation dose. A prompt necking at yield (with little uniform ductility) was observed after irradiation at 300 °C but no embrittlement was observed up to 10 dpa. Ductilization by irradiation–the radiation-induced increase of ductility–was observed for the 600 °C irradiation only and it occurred in low dose range only. The neutron irradiation increased the data variation in many tensile property datasets, particularly after 600 °C irradiation, and no clear dependence of tensile properties on build thickness or sampling location was observed.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructural Simulation of Ion-Irradiated Natural Rocks and Minerals

This report details the contribution of Oak Ridge National Laboratory to the Nuclear Energy University Program (NEUP) project, "Rapid Characterization of Concrete Mineralogy using Multi-Scale Tools" led by the University of Illinois at Urbana-Champaign. To support the subsequent license renewal (SLR) of the US nuclear power plants (NPPs) fleet, the performance of concrete-forming aggregates against neutron irradiation needs to be assessed. Ion irradiation is proposed as a rapid, cost-effective surrogate method to full neutron irradiation testing. To complement the characterization of ion-irradiated rock specimens, ORNL has run finite-elements and fast-Fourier transform (FFT-based) simulations using the codes MARS and Microstructure-Oriented Scientific Analysis of Irradiated Concrete (MOSAIC). The main conclusion of this analysis is that the apparent post–ion-irradiation step-height underestimates the accumulated free radiation-induced volumetric expansion (RIVE) in the ion-implanted depth by about 15% at full amorphization and about 25% toward the beginning of the ion irradiation experiment. The discrepancy is explained by the fact that the step height is proportional to the sum of the RIVE (i.e., one third of the RIVE) and the irradiation-assisted plastic strains in the vertical direction (lower than two thirds of the RIVE). Because of the large lateral compressive stresses caused by the restraining effect of the unirradiated substrate, the stress field in the mineral grains and at the grain boundary (GB) in the ion-implanted layer is much different than the comparable stress field occurring during neutron irradiation. Hence, the mismatch RIVE causing cracks in the rock-forming minerals leads to different cracking patterns. Ion irradiation continues to be used as a rapid technique to assess the RIVE potential of rocks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Status report on HFIR irradiation of optimized alumina forming alloys

Properties of FeCrAl alloys under neutron irradiation are of interest because of these materials’ potential application as accident-tolerant fuel cladding in nuclear systems. In parallel, alumina-forming austenitic (AFA) alloys are of interest for use as structural materials in advanced nuclear systems for their potential higher resistance to embrittlement and high-temperature steam oxidation resistance. An irradiation campaign for fiscal year 2024 has been developed under the Advanced Fuels Campaign to perform irradiation testing of various FeCrAl and AFA alloys in Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR). The goals of this irradiation campaign are to (1) study the impact of minor alloying elements on the neutron-irradiated mechanical properties of FeCrAl alloys and (2) collect neutron-irradiated mechanical properties on AFA alloys for comparison with those of FeCrAl alloys. This campaign will include both tensile and fracture toughness specimens tested following HFIR irradiation at temperatures representative of normal operating conditions in light-water reactors. The pre-irradiation characterization to date, the irradiation plan for the FeCrAl and AFA specimens, and the subsequent post-irradiation experimental test plan are presented in this report, along with the status of HFIR builds and scheduled insertion dates.

36 MATERIALS SCIENCE↗