SEARCH · Engineering Papers

Results for “IRRADIATION”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 505 records · Page 28

All sources NASA NTRS DOE OSTI

DOE OSTI2023

Small Scale Creep Testing of 14YWT via In-situ Transmission Electron Microscopy Irradiation and Nanoindentation

The next generation of nuclear materials must withstand harsh operating conditions such as high temperatures and irradiation doses. Nanostructured ferritic alloys like oxide dispersion strengthened steels, especially 14YWT, have shown promise as a structural material to withstand these harsh operating conditions. For application, understanding the irradiation enhanced creep of the structural components is fundamental to evaluating the service life in a reactor. Challenges with evaluating irradiation enhanced creep are related to the limited space in nuclear reactors and the expense of conducting post irradiation experiments on irradiated cladding. Ion irradiations are considered to expedite irradiation testing, but such experiments have restricted depth of penetration into the material, limiting the ability to characterize changes to material properties after irradiation. Small scale mechanical testing can be used with ion beam irradiations as a potential avenue to evaluate the irradiation enhanced creep of materials. In this study, in-situ transmission electron microscopy nanopillar creep studies on 14YWT were performed and simultaneously ion beam irradiated with 2.8 MeV Au4+ ions. It was observed that the ion beam irradiation did increase the measured strain rate of the materials. In addition, ex-situ nanoindentation creep studies were performed over a range of temperatures on control 14YWT, and it was observed that there was a change in the deformation mechanism between 873 K and 1073 K that agrees well with macro-scale mechanical testing. Furthermore, these results show promise for applying these techniques to nuclear materials in the future.

14 YWT↗

DOE OSTI2018

Effects of polarized and non-polarized red-light irradiation on proliferation of human Wharton's jelly-derived mesenchymal cells

Light emitting diode (LED) irradiation has recently been introduced as an encouraging strategy for promotion of cell proliferation. Human umbilical cord Wharton's jelly-derived mesenchymal (hUCM) cells are among the most available mesenchymal cells with a promising application in regenerative medicine. The aim of the present study was to examine the effect of polarized (PL) and non-polarized (NPL) red-light emitted by LED on various proliferation properties of hUCM cells. Cell proliferation was assessed 48 h after irradiation of hUCM cells by different energy densities. Cell density increased to a significant level both in PL and NPL irradiation at 0.954 J/cm{sup 2} following WST-1 assay. Staining of irradiated and non-irradiated cells with Hoechst after 3 and 6 days revealed an increased proliferation rate in irradiated cells, but the non-irradiated cells proliferated more than irradiated cells at day 9 of cultivation. Similar results were obtained in trypan blue assay. Scratch repair test for 18 h with an interval of 6 h did not reveal a significant difference between irradiated and non-irradiated cells. In addition, CFU-F assay in PL irradiated cells was higher than control when 500 cells/plate was cultivated. Totally, this study revealed that hUCM cells could be induced to achieve higher number of cells by PL and NPL red-light irradiation after 48 h.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2016

BNL Irradiation Damage Studies of the Metal Matrix Composite Mo-GR Considered for High Luminosity LHC Collimator Upgrade

A comprehensive experimental/analytical effort on the resilience of Mo-GR metal matrix composite to high energy proton beam irradiation combined with elevated temperatures has been undertaken at BNL on behalf of the CERN LHC Collimation effort as part of an effort to identify novel materials that will exhibit optimal performance in the high luminosity LHC upgrade. While the Mo-Gr composite was one of the four (4) different materials or composites (Mo, Glidcop, Cu-CD and Mo-Gr) that constituted the High Luminosity LHC collimation material campaign, this report focuses solely on Mo-GR with some reference to its constituents Mo and graphite (graphite studied in previous research or irradiation campaigns along with the carbonfiber composites). The multi-faceted study on Mo-GR (one of several different alloys and matrices) performed at BNL included the following stages: Irradiation damage phase using the 200 MeV proton beams of the BNL Linac and the BLIP target beamline. Specifically, Mo-GR samples encapsulated under vacuum in special capsules were exposed to direct protons as part of a multi-capsule, multi-material irradiation array. The long-term irradiation of ~8 weeks resulted in an accumulated fluence at the Mo-GR of ~10 21 p/cm 2 . The irradiation was conducted in two sessions (RUN 2013 and RUN 2014 of the BLIP operations in tandem with isotope production). Irradiation damage experiments using spallation neutrons at the BLIP beamline generated by 118 MeV protons that interact with and are stopped by an array of isotope producing targets. The Mo-GR capsule combined the effects of short-term proton irradiation and the effects of the mixed irradiating spectrum. Spallation-induced irradiation of Mo-GR at sub-zero temperatures utilizing the mixed field generated by spallation of 28 MeV protons of the BN Tandem on Mo targets Gamma spectrum analysis of the heavily irradiated Mo-Gr to enable the correlation of predicted isotope profile from Mo-Gr by transport codes and therefore the utilization of the validated transport code in other aspects of the damage induced on Mo-Gr by irradiation.

43 PARTICLE ACCELERATORS↗

DOE OSTI2020

First Examination of Irradiated Fuel with Pulsed Neutrons at LANSCE (Preliminary Results)

We present preliminary results on the characterization of an irradiated U-lOZr-lPd fuel sample that was prepared from the irradiated AFC-3A-R5A sample. U-lOZr metallic fuels are researched as host materials for potential transmutation fuels and the addition of palladium strives to bind lanthanides, thus preventing fuel-cladding chemical interactions (FCCI). These interactions limit the lifetime of metallic fuels and are caused by migration of lanthanide fission products to the periphery of the fuel slug, where they start to interact with the D9 or HT9 steel cladding, ultimately leading to failure of the mechanical integrity of the cladding. Neutrons offer bulk characterization of irradiated materials for which X-ray tomography methods are not suitable due to the immense gamma background emitted from the samples. In particular pulsed neutrons provide information from the ability to resolve the neutron energy using their time-of-flight and thus the potential to utilize neutron absorption resonance to characterize the spatial distribution of isotopes. This, in turn, may allow to characterize the distribution of fission and neutron capture products non-destructively and may ultimately be applied to the bulk of an irradiation capsule prior to destructive post-irradiation examination to identify regions of interest. To allow the characterization of entire irradiation capsules, a cask is under development in the advanced post-irradiation work package at LANL and progress on this development was reported elsewhere. In parallel, an irradiated U-lOZr-lPd sample cut from the AFC-3AR5A irradiation was shipped to LANL and will be fully characterized with an NSUF funded rapid turnaround experiment (RTE) in the 2020 LANSCE run cycle. The sample emits at a dose rate of ~3R/hr on contact and is therefore manageable with remote handling, without requiring a cask. The disk-shaped material is larger than samples prepared for analysis using electron or X-ray methods and is therefore an intermediate step towards characterization of bulk samples at LANSCE. However, since it covers the full diameter of the irradiated fuel slug, some insight on redistribution of elements, spatially resolved information on microstructure, e.g. phase composition and texture, will be possible using the pulsed neutron-based methods developed for fuel characterization at LANSCE. This report describes the development of procedures to handle the sample at LANSCE as well as preliminary data and results from tests conducted in December 2019 on the energy-resolved neutron imaging (ERNI) beam line at flight path 5 and the high pressure-preferred orientation diffractometer (HIPPO) at LANSCE. This effort is a collaboration between LANL, INL, and ORNL. To compare our capabilities with prior work, we present an overview of previously reported bulk characterization of irradiated or spent fuels. The overview addresses neutron diffraction and neutron absorption resonance spectroscopy, both of which have only few reported applications on irradiated or spent nuclear fuel, as well as neutron radiography. This literature review was already described in a previous report but is repeated here to put our current efforts in context of previous work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Irradiation of Advanced Cladding Specimens in the High Flux Isotope Reactor: Capsule Designs and Test Matrix

The Advanced Fuels Campaign (AFC) has initiated the Advanced Reactor Cladding (ARC) irradiation campaign to generate irradiation performance data for candidate fuel cladding concepts. The campaign includes a diverse set of ferritic/martensitic steels, oxide dispersion strengthened (ODS) alloys, FeCrAlbased alloys, coated materials, and welded cladding specimens produced through multiple US Department of Energy (DOE) programs and international collaborations. Three complementary experimental thrusts comprise the campaign: tensile testing (ARC Tensile) to rapidly screen candidate alloys, fracture toughness testing (ARC Fracture) to evaluate irradiation effects on crack resistance, and tubular weld testing (ARC Weld) to quantify irradiation-induced changes in the mechanical performance of end cap welds. This report documents the irradiation campaign design, including the selected materials, specimen types, irradiation matrix, and capsule designs for irradiation within the High Flux Isotope Reactor (HFIR). A total of 14 irradiation capsules were developed to achieve target irradiation temperatures between 300°C and 600°C and doses up to 30 dpa. Thermal analyses were performed using finite element methods to establish capsule geometries capable of achieving the desired specimen temperatures while accommodating differences in specimen geometry and material properties. The resulting capsule designs provide the basis for irradiation of the AFC-ARC experimental matrix and subsequent post-irradiation examination to assess the effects of neutron irradiation on advanced cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Comparison of Ion and Neutron Irradiations to 3 dpa at 500C in Ferritic-Martensitic Alloys

The growing global demand for energy will increasingly call upon advanced nuclear fission reactors to supply safe and reliable electricity. The structural and fuel cladding components of these reactors will be subject to extreme conditions of irradiation damage up to several hundred displacements per atom (dpa) at temperatures as high as 700°C. Ferritic-martensitic (F-M) steels are leading candidates for these challenging conditions due to their strength and dimensional stability under irradiation. In order to accelerate the process for evaluating F-M alloys, charged particles are increasingly being used to emulate neutron irradiations. Charged particle irradiations allow the possibility of conducting irradiation experiments within a shorter time period (i.e. at a rate up to 4 orders of magnitude faster) and with minimal radioactivation of the material, enabling lower cost and faster turnaround of post irradiation examination and analysis. However, the irradiation dose rate, damage cascade morphologies, and irradiation damage depth profiles all differ widely between protons, heavier ions, and neutrons. Currently, there is limited understanding of the significance of these physical differences and how they manifest in the irradiated microstructure and mechanical properties of F-M steels. The objective of this study is to evaluate charged particles as a surrogate for neutron irradiations in F-M alloys by assessing common irradiation conditions using Fe++ ions, protons, and neutrons. Keeping the temperature and dose consistent enables isolation of the effects of each irradiating particle and their respective dose rates and cascade morphologies.

Swenson, M.J.↗

DOE OSTI2016

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Atom probe tomography characterization of ion and neutron irradiated Alloy 800H

Alloy 800H is a high temperature and creep resistant alloy, and is considered as a candidate alloy for use in Generation IV nuclear reactor systems. To clarify the alloy's behavior under irradiation, samples were subjected to single ion irradiations up to 20 dpa at 440°C, a dual ion/He irradiation to 17 dpa at 460°C, and a neutron irradiation to 17 dpa at 385°C. The irradiated microstructures were characterized using atom probe tomography to complement previously published transmission electron microscopy studies. Additionally, after single ion irradiation, sparse fine Al and Ti clusters were observed after 1dpa, while high number densities of nanoscale Ni-Al-Ti clusters, Cr-Ti rich carbides, and Si-decorated dislocation loops developed after 10 and 20 dpa. The microstructure formed during single ion irradiation exhibited a strong depth dependence. No significant differences were observed between the single and dual ion irradiations. While Ni-Al-Ti clusters, Cr-Ti rich carbides, and Si-decorated dislocation loops were also observed after neutron irradiation, the neutron-irradiated microstructure differed from those found in the ion-irradiated samples.

36 MATERIALS SCIENCE↗

DOE OSTI2023

High-temperature steam oxidation study of irradiated FeCrAl defueled specimens

Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO 2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO 2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. Here, the ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Status of Triso fuel irradiations in the advanced test reactor supporting high-temperature gas-cooled reactor designs - 18468

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven experiments on low-enriched uranium tri-structural isotropic (TRISO) particle fuel. Irradiations and fuel development are being accomplished at Idaho National Laboratory s Advanced Test Reactor to support development of the next generation of reactors in the United States. The experiments are being irradiated over several years to demonstrate and qualify new TRISO-coated particle fuel for use in high-temperature gas reactors, with goals of providing irradiation performance data to support fuel process development, qualifying fuel for normal operating conditions, supporting development and validation of fuel performance and fission product transport models and codes, and providing irradiated fuel and materials for post-irradiation examination and safety testing. The experiments are irradiated in an inert sweep gas atmosphere, with individual online temperature monitoring and control of each capsule and online fission product monitoring of effluent to track individual capsule fuel performance. The first experiment (AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Because the purpose of this experiment was to provide data on fission product migration and retention in the next generation reactor, the design of this experiment was significantly different from the first two experiments. The final experiment (AGR-5/6/7) is scheduled to begin irradiation in early summer 2017. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Effect of proton irradiation on anatase TiO 2 nanotube anodes for lithium-ion batteries

The role of defects in the charge transfer and transport properties of electrode materials for lithium-ion batteries has recently garnered increased interest. It is widely known that ion irradiation promotes the formation of defects within a crystalline solid. Within all ion species used for irradiation, protons are expected to create primarily simple Frenkel pair point defects without significantly changing the stoichiometry of the damaged region of the target material. This work investigates the effect of proton irradiation at varying temperatures on the electrochemical properties of anatase TiO 2 nanotube (TiO2-NT) electrode for lithium-ion battery applications. Anatase TiO 2 -NTs are irradiated at both room temperature (25 °C) and 250 °C and compared with non-irradiated control specimens. Characterization by Raman spectroscopy and XRD suggests that the irradiation at both temperatures does not alter the long-range order of the nanotubes. Yet, high-resolution TEM reveals that defect clusters are formed upon irradiation and increase in size with increasing temperature. Both irradiated samples exhibit increased capacity and enhanced rate capability compared with the non-irradiated control, which can be explained by increased storage sites as well as improved Li + diffusivity due to the presence of irradiation-induced defects. This study presents a unique perspective on pathways to engineer functional nanostructured electrode materials by tailoring irradiation conditions.

25 ENERGY STORAGE↗

DOE OSTI2020

Structural damage and phase stability of Al 0.3 CoCrFeNi high entropy alloy under high temperature ion irradiation

In this study, an initially single phase high entropy alloy (HEA) Al 0.3 CoCrFeNi was irradiated by 3 MeV Au ions to a fluence of 6 × 10 15 cm –2 (~31 dpa at damage peak) at four different temperatures ranging from 250 °C to 650 °C. Transmission electron microscopy (TEM) and Atom probe tomography (APT) were employed to study the evolution of structural damage and phase stability with irradiation temperature. Al 0.3 CoCrFeNi exhibited a similar evolution of irradiation-induced defects with temperature as compared with conventional FCC alloys. At 250 °C and 350 °C, most of the visible irradiation-induced defects were faulted 1/3$\langle{111}\rangle$ dislocation loops. As the irradiation temperature increased to 500 °C, perfect 1/2$\langle{110}\rangle$ dislocation loops were observed along with the faulted loops. At the highest irradiation temperature 650 °C, only dislocation lines and networks could be observed. Regarding phase stability, the 3 MeV Au irradiation was observed to suppress the precipitation of (Ni, Al)-enriched nano clusters and the L12 ordered structure at irradiation temperatures 250 °C to 500 °C whereas precipitation of the B2 ordered structure was accelerated at 650 °C. This resulted in qualitatively opposite precipitation behavior between the ion irradiated damage region and unirradiated region at 500 °C and 650 °C. The opposite phase stability of the ion-irradiated damage region and unirradiated region at different temperatures is attributed to the competing effects of ballistic dissolution vs irradiation enhanced diffusion on precipitation.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Irradiated fibroblasts promote epithelial–mesenchymal transition and HDGF expression of esophageal squamous cell carcinoma

Recent evidence suggested that nonirradiated cancer-associated fibroblasts (CAFs) promoted aggressive phenotypes of cancer cells through epithelial–mesenchymal transition (EMT). Hepatoma-derived growth factor (HDGF) is a radiosensitive gene of esophageal squamous cell carcinoma (ESCC). This study aimed to investigate the effect of irradiated fibroblasts on EMT and HDGF expression of ESCC. Our study demonstrated that coculture with nonirradiated fibroblasts significantly increased the invasive ability of ESCC cells and the increased invasiveness was further accelerated when they were cocultured with irradiated fibroblasts. Scattering of ESCC cells was also accelerated by the supernatant from irradiated fibroblasts. Exposure of ESCC cells to supernatant from irradiated fibroblasts resulted in decreased E-cadherin, increased vimentin in vitro and β-catenin was demonstrated to localize to the nucleus in tumor cells with irradiated fibroblasts in vivo models. The expression of HDGF and β-catenin were increased in both fibroblasts and ESCC cells of irradiated group in vitro and in vivo models. Interestingly, the tumor cells adjoining the stromal fibroblasts displayed strong nuclear HDGF immunoreactivity, which suggested the occurrence of a paracrine effect of fibroblasts on HDGF expression. These data suggested that irradiated fibroblasts promoted invasion, growth, EMT and HDGF expression of ESCC. - Highlights: • Irradiated CAFs accelerated invasiveness and scattering of ESCC cell lines. • Irradiated CAFs promoted EMT of ESCC cells. • Irradiated fibroblasts induced nuclear β-catenin relocalization in ESCC cells. • Irradiated fibroblasts increased HDGF expression in vitro and in vivo.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2014

Deficiency in Homologous Recombination Renders Mammalian Cells More Sensitive to Proton Versus Photon Irradiation

Purpose: To investigate the impact of the 2 major DNA repair machineries on cellular survival in response to irradiation with the 2 types of ionizing radiation. Methods and Materials: The DNA repair and cell survival endpoints in wild-type, homologous recombination (HR)-deficient, and nonhomologous end-joining-deficient cells were analyzed after irradiation with clinically relevant, low-linear energy transfer (LET) protons and 200-keV photons. Results: All cell lines were more sensitive to proton irradiation compared with photon irradiation, despite no differences in the induction of DNA breaks. Interestingly, HR-deficient cells and wild-type cells with small interfering RNA-down-regulated Rad51 were markedly hypersensitive to proton irradiation, resulting in an increased relative biological effectiveness in comparison with the relative biological effectiveness determined in wild-type cells. In contrast, lack of nonhomologous end-joining did not result in hypersensitivity toward proton irradiation. Repair kinetics of DNA damage in wild-type cells were equal after both types of irradiation, although proton irradiation resulted in more lethal chromosomal aberrations. Finally, repair kinetics in HR-deficient cells were significantly delayed after proton irradiation, with elevated amounts of residual γH2AX foci after irradiation. Conclusion: Our data indicate a differential quality of DNA damage by proton versus photon irradiation, with a specific requirement for homologous recombination for DNA repair and enhanced cell survival. This has potential relevance for clinical stratification of patients carrying mutations in the DNA damage response pathways.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2016

Decreasing Irradiated Rat Lung Volume Changes Dose-Limiting Toxicity From Early to Late Effects

Purpose: Technological developments in radiation therapy result in smaller irradiated volumes of normal tissue. Because the risk of radiation therapy-induced toxicity generally depends on irradiated volume, changing volume could change the dose-limiting toxicity of a treatment. Recently, in our rat model, we found that early radiation-induced lung dysfunction (RILD) was closely related to irradiated volume dependent vascular remodeling besides inflammation. The exact relationship between early and late RILD is still unknown. Therefore, in this preclinical study we investigated the dose-volume relationship of late RILD, assessed its dependence on early and late pathologies and studied if decreasing irradiated volume changed the dose-limiting toxicity. Methods and Materials: A volume of 25%, 32%, 50%, 63%, 88%, or 100% of the rat lung was irradiated using protons. Until 26 weeks after irradiation, respiratory rates were measured. Macrovascular remodeling, pulmonary inflammation, and fibrosis were assessed at 26 weeks after irradiation. For all endpoints dose-volume response curves were made. These results were compared to our previously published early lung effects. Results: Early vascular remodeling and inflammation correlated significantly with early RILD. Late RILD correlated with inflammation and fibrosis, but not with vascular remodeling. In contrast to the early effects, late vascular remodeling, inflammation and fibrosis showed a primarily dose but not volume dependence. Comparison of respiratory rate increases early and late after irradiation for the different dose-distributions indicated that with decreasing irradiated volumes, the dose-limiting toxicity changed from early to late RILD. Conclusions: In our rat model, different pathologies underlie early and late RILD with different dose-volume dependencies. Consequently, the dose-limiting toxicity changed from early to late dysfunction when the irradiated volume was reduced. In patients, early and late RILD are also due to different pathologies. As such, new radiation techniques reducing irradiated volume might change the dose-limiting toxicity of the radiation therapy treatment.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2020

Temperature effects on damage evolution in ion-irradiated NiCoCr concentrated solid-solution alloy

In this work, the damage evolution in equiatomic NiCoCr single crystals irradiated at 138 and 500 K with 1.5 MeV Ni+ ions has been investigated using ion channeling analysis and compared with previous results obtained at 300 K under identical irradiation conditions. As expected, irradiation temperature has a major influence on damage evolution and long-range effects in ion-irradiated NiCoCr. A change in relative irradiation resistance as a function of temperature is revealed when the ion channeling data are compared with previous ion channeling results from Ni-irradiated NiFe at 150, 300 and 500 K. Additionally, NiCoCr has superior irradiation resistance than NiFe under ion irradiation at ≤ 300 K; surprisingly, the relative irradiation resistance is reversed under ion irradiation performed at > 300 K (i.e., 500 K), which may be attributed to local chemical ordering. These observations indicate that the NiFe alloy exhibits a more favorable chemical environment for suppressing deleterious microstructural degradation under high temperature irradiation than the NiCoCr alloy.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe 3+ and 3.42 MeV energy degraded He 2+ ) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. Furthermore, the comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Heavy ion irradiation-induced microstructural evolution in pyrochlore Lu{sub 2}Ti{sub 2}O{sub 7} at room temperature and 723 K

Polycrystalline pyrochlore Lu{sub 2}Ti{sub 2}O{sub 7} pellets were irradiated with 600 keV Kr{sup 3+} at room temperature and 723 K to a fluence of 4×10{sup 15} ions/cm{sup 2}, corresponding to an average ballistic damage dose of 10 displacements per atom in the peak damage region. Irradiation-induced microstructural evolution was examined by grazing incidence X-ray diffraction, and cross-sectional transmission electron microscopy. Incomplete amorphization was observed in the sample irradiated at room temperature due to the formation of nano-crystal which has the identical structure of pyrochlore, and the formation of nano-crystal is attributed to the mechanism of epitaxial recrystallization. However, an ordered pyrochlore phase to a swelling disordered fluorite phase transformation is occurred for the Lu{sub 2}Ti{sub 2}O{sub 7} sample irradiated at 723 K, which is due to the disordering of metal cations and anion vacancies. - Graphical Abstract: Polycrystalline pyrochlore Lu{sub 2}Ti{sub 2}O{sub 7} pellets were irradiated with 600 keV Kr{sup 3+} to a fluence of 4×10{sup 15} ions/cm{sup 2} at room temperature and 723 K, Incomplete amorphization was observed in the sample irradiated at room temperature due to the formation of nano-crystal. However, an ordered pyrochlore phase to a swelling disordered fluorite phase transformation is occurred for the Lu{sub 2}Ti{sub 2}O{sub 7} sample irradiated at 723 K, which is due to the disordering of metal cations and anion vacancies. - Highlights: Pyrochlore Lu{sub 2}Ti{sub 2}O{sub 7} pellets were irradiated by heavy ions at RT and 723 K. At RT irradiation, ~75% of amorphization was achieved. The nano-crystals were formed in the damage layer at RT irradiation. The formed nano-crystals enhanced the radiation tolerance of Lu{sub 2}Ti{sub 2}O{sub 7}. A pyrochlore to fluorite phase transformation was observed at 723 K irradiation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗