Suppressing irradiation induced grain growth and defect accumulation in nanocrystalline tungsten through grain boundary doping
Not Available
Cunningham, W. Streit (ORCID:0000000331663563)↗
SEARCH · Engineering Papers
Results for “Irradiation growth”
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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The criteria in above-bandgap photo-irradiation in molecular beam epitaxy growth of heterostructure of dissimilar growth temperature
Above-bandgap photo-irradiation is known to improve the low temperature growth of II-VI semiconductors, but the trade-offs in the substrate temperature and light source power density are not well known. We investigated these effects on the growth of ZnSe epilayers on GaAs. Here, we find that the above-bandgap photo-irradiation can improve the ZnSe epilayer without substantially negatively impacting the underlying GaAs epilayer only if the laser energy is below a threshold intensity. When the threshold is exceeded, the growth rate drops, the optical properties of ZnSe layer deteriorate and interface intermixing is enhanced. Together, cross-sectional transmission electron microscopy, energy dispersive spectroscopy and photoluminescence results suggest that photo-irradiation at moderate to high laser energies produces a trade-off in interface intermixing and planar defect formation. Most importantly, the damage produced by high laser energies does not start at the interface but instead in the bulk. Further flexibility for selecting the temperature and photo-irradiation intensities could be realized by turning on the laser irradiation after the ZnSe growth has been initiated, limiting the potential intermixing at the interface.
36 MATERIALS SCIENCE↗
Effect of Water Chemistry on Crack Growth Rate in Neutron Irradiated X-750 and XM-19
The Advanced Test Reactor (ATR) National Science User Facility (NSUF) was used to study the effect of neutron damage on crack growth rates (CGR) under simulated boiling water reactor (BWR) water chemistry conditions using a fast neutron spectrum to accelerate the damage process. Nickel based Alloy X-750 and XM-19 were irradiated in the ATR center flux trap (CFT), with this experiment marking the first civilian project to utilize this position within the reactor. 0.4T-CT specimens were used to study crack growth rates for two sample fluences (~1.9 x 1020 n/cm2 and ~9.5 x 1020 n/cm2, E > 1MeV)) in both normal water chemistry (NWC) at 2.5 ppm O2 and hydrogen water chemistry (HWC) at ~80 ppb H2. CGR testing was conducted at a temperature of 288 °C. A decrease in CGR was observed for both alloys with the average ratio (CGR in NWC/CGR in HWC) being a factor of 281 for the XM-19 samples and a factor of 27.8 for the X-750 samples. Observed CGRs in this study showed little difference compared to non-irradiated samples (the non-irradiated XM-19 was 20% cold worked), suggesting limited sensitivity of these susceptible alloys to irradiation effects. Detailed results of CGR response to varying environmental electrochemical potential (ECP) and water purity will be presented in the context of crack growth mitigation in different water chemistries and the change in material behavior relative to non-irradiated samples of the material.
36 MATERIALS SCIENCE↗
Effect of Water Chemistry on Crack Growth Rates in Neutron Irradiated X-750 and XM-19
The Advanced Test Reactor (ATR) National Science User Facility (NSUF) was used to study the effect of neutron damage on crack growth rates (CGR) under simulated boiling water reactor (BWR) water chemistry conditions using a fast neutron spectrum to accelerate the damage process. Nickel based Alloy X-750 and XM-19 were irradiated in the ATR center flux trap (CFT), with this experiment marking the first civilian project to utilize this position within the reactor. 0.4T-CT specimens were used to study crack growth rates for two sample fluences (~1.9 x 1020 n/cm2 and ~9.5 x 1020 n/cm2, E > 1MeV) in both normal water chemistry (NWC) at 2.5 ppm O2 and hydrogen water chemistry (HWC) at ~80 ppb H2. A decrease in CGR was observed with HWC for both alloys with the average ratio (CGR in NWC/CGR in HWC) being a factor of 281 for the XM-19 samples and a factor of 27.8 for the X-750 samples. Observed CGRs in this study showed little difference compared to non-irradiated samples (the non-irradiated XM-19 was 20% cold worked), suggesting limited sensitivity of these susceptible alloys to irradiation effects. The CGR response to varying environmental electrochemical potential (ECP) and water purity will be presented.
36 MATERIALS SCIENCE↗
Utility of DF-1 for Radioprotection in Lymphocytes
The development of degenerative changes in the vasculature, such as atherosclerosis, is a known consequence of exposure to ionizing radiation, and is thus a concern for astronaut health following long duration space flight. Cellular damage caused by radiation is due to free radical generation and DNA damage. The goal of this project was to assess the ability of a C60-derivative, DF-1, to mitigate cellular damage resulting from radiation exposure in primary human lymphocytes. DF-1 is a water-soluble C60 fullerene encapsulated in dendrimeric functional groups that is proposed to exhibit antioxidant properties. Human lymphocytes are radiosensitive and travel throughout the body potentially causing bystander effects in any tissues they contact. These cells were subjected to varying doses of gamma radiation in the presence or absence of DF-1. Cells were collected at 48 hours post-irradiation for chromosomal aberration studies and at 72 hours post-irradiation for micronuclei studies. These studies showed that the irradiated cells contained more chromosomal aberrations and micronuclei than the control cells. Addition of the DF-1 reduced the amount of observed DNA damage in the irradiated cells. Growth curves were measured for the lymphocytes exposed to 0 and 4 Gray gamma irradiations, and we observed less growth in the cells irradiated at 4 Gy. 2,7-dichlorofluorescein diacetate was used to detect reactive oxygen species production, and increased production of ROS was observed in the irradiated lymphocytes. Human lymphocytes were subjected to varying doses of gamma or photon radiation in the presence and absence of DF-1 and a known radioprotectant, amifostine. After irradiation, the production of reactive oxygen species, growth curves and cell viability were measured. These cells were also collected to quantify chromosomal aberrations and micronuclei formation. We predict that irradiated cells will show the most damage and that DF-1 will provide protective effects similar to those of amifostine, an established radioprotectant.
Reynolds, Julia↗
In Situ Grain Growth of Nanograined Magnetite under Ion Irradiation at Room Temperature and 500 ℃
Grain growth studies of nanograined metals under both thermal and ion irradiation conditions has been studied extensively. Oxides, on the other hand, have not been studied as extensively, especially under irradiation. Some studies have focused on nanograined oxide grain growth on ZnO under thermal conditions only and on SnO 2 and CeO 2 using 250 keV Ar 2+ and 3 MeV Au + ions at 300 K, respectively. There does not appear to be any studies on grain growth of nanosized oxides which are commonly found to form under light water reactor conditions. Oxides such as magnetite (Fe 3 O 4 ), hematite (α-Fe 2 O 3 ), maghemite (γ-Fe 2 O 3 ), and wustite (FeO) are examples of common oxides which have not been studied. Here, this study focused on the grain growth of nanograined Fe 3 O 4 under both thermal and ion irradiation conditions in situ inside a Transmission Electron Microscope (TEM) to better understand the kinetics of grain growth of oxides.
36 MATERIALS SCIENCE↗
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↗
Abnormal grain growth driven by high-temperature proton irradiation in nanocrystalline Ni
In this research we examine a nanocrystalline Ni thin film exposed to high-temperature proton irradiation and compare it with as-deposited and annealed-only counterparts. Despite lacking thermal spikes typical of heavy ions, 400 °C proton irradiation drives pronounced grain growth in select grains, whereas annealing alone yields only modest coarsening. Grain-boundary statistics show fewer low-angle boundaries (10–20°) and more high-angle boundaries (55–60°), consistent with irradiation-enhanced mobility of high-misorientation boundaries. The irradiated films retain a random texture, with no evidence of texture development or sharpening. Mechanisms, such as radiation-enhanced grain boundary diffusion, beam-induced heating, and ion channeling-mediated selective grain growth, are unlikely to be the predominant drivers to explain the resultant microstructure. Instead, we suggest irradiation-induced modifications of grain-boundary structure, including possible complexion transitions, as one plausible explanation for this selective grain growth and retention of random texture. However, additional temperature–dose studies are required to confirm the mechanism.
36 MATERIALS SCIENCE↗
Coupled cluster and dislocation dynamics modeling of microstructure evolution in irradiated materials
We develop here a coupled cluster and dislocation dynamics framework to study the microstructure evolution of irradiated materials. The framework not only accounts for the three dimensional diffusion of radiation-generated clusters, but also their interaction with dislocation networks and the resultant climb motion of discrete dislocations within finite crystals. The framework is solved with a superposition solution scheme, and is applied to investigate the evolution of the irradiation-induced dislocation loops in zirconium (Zr), considering the effects of various bias factors including the diffusion anisotropy difference (DAD) of interstitials and interstitial clusters, the dislocation bias of defects to discrete dislocation segments, and the production bias of defects from the radiation cascade. We find that the DAD is the most critical factor influencing the kinetics of the loop evolution in Zr, while the recombination/interaction of mobile defects can induce a strong spatial dependence of the loop evolution together with the DAD. Here, the method is also adopted to study the evolution of interstitial $\langle$a$\rangle$ and vacancy $\langle$c$\rangle$ dislocation loop ensembles consistent with the microstructure observed during irradiation-induced growth of Zr. Our findings not only reveal the spatial dependence of the size and ellipticity of the dislocation loops, but also suggest a limit on the anisotropy factor of interstitials to reproduce the co-growth of $\langle$a$\rangle$ and $\langle$c$\rangle$ loops in zirconium, in good agreement with experimental observations and other simulation results.
Bias factors↗
Positron Emission Tomography Imaging of Functional Transforming Growth Factor β (TGFβ) Activity and Benefit of TGFβ Inhibition in Irradiated Intracranial Tumors
Transforming growth factor β (TGFβ) promotes cell survival by endorsing DNA damage repair and mediates an immunosuppressive tumor microenvironment. Thus, TGFβ activation in response to radiation therapy is potentially targetable because it opposes therapeutic control. Strategies to assess this potential in the clinic are needed.
62 RADIOLOGY AND NUCLEAR MEDICINE↗
Does solar radiation affect the growth of tomato seeds relative to their environment?
The purpose of this experiment is to sequentially study and analyze the data collected from the germination and growth of irradiated Rutgers Supreme tomato seeds to adult producing plants. This experiment will not use irradiated seeds as a control as I plan to note growth in artificial verses natural environment as the basic experiment.
Holzer, Kristi↗
A Review of Irradiation Damage and Effects in α-Uranium
Understanding irradiation damage and effects in α-uranium (α-U) is critical to modeling the behavior of U-based metallic fuels. The aim of this review is to address the renewed interest in U-based metallic fuels by examining the state-of-the-art knowledge associated with the effect of irradiation on the microstructure, dimensional changes, and properties of α-U. We critically review the research progress on irradiation-induced growth and swelling, the enhancement of plastic flow and superplasticity by irradiation, and the effect of irradiation on thermal and electrical properties of α-U. Finally, we outline the research directions that require advancements, specifically the need to carry out fundamental research on several of the less understood mechanisms of irradiation damage and effects in α-U.
36 MATERIALS SCIENCE↗
A photosensitive sustainable lignin nanoplatform for multimodal image-guided mitochondria-targeted photodynamic and photothermal therapy
Photoactivated nanocarriers exhibit significant potential for anticancer therapy, but complex design strategies, unsustainable substrates, and short wavelengths limit their practical application. Here, we designed a new lignin-derived photoactivated nanomaterial that exploits the sensitivity of the β-O-4 bond of lignin to singlet oxygen. This sustainable product was loaded with mitochondria-targeting chlorin e6 and black phosphorus quantum dots (BPQDs) to produce BPQDs@N-LgC NPs, which were used for mitochondria-targeted fluorescence/photoacoustic-guided photothermal and photodynamic therapy. When irradiated at 808 nm, the BPQDs in BPQDs@N-LgC NPs exhibited good photothermal conversion, which allowed photoacoustic imaging and inhibited tumor growth. When irradiated at 660 nm, the BPQDs@N-LgC NPs generated fluorescence and reactive oxygen species, which allowed photoluminescence imaging and further inhibited tumor growth. Cleavage of the β-O-4 bond of lignin by photo-triggered reactive oxygen species degraded the NPs and released the BPQDs, facilitating rapid excretion of the therapeutic nanomaterials. Our rationally designed BPQDs@N-LgC NPs exhibited good therapeutic efficacy, both in vitro and in vivo.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Carbon dioxide effects on potato growth under different photoperiods and irradiance
The effects of atmospheric CO2 concentration, photosynthetic photon flux (PPF), and the length of the photoperiod on the tuber yield were investigated for three potato cultivars (Norland, Russet Burbank, and Denali), by growing these cultivars for 90 days in atmospheres containing 350 or 1000 micromol/mol CO2, at photoperiods of 12- or 24-hr, and at PPFs of 400 or 800 micromol/sq m per sec. Air temperatures and relative humidity were kept at 16 C and 70 percent, respectively. It was found that the tuber yield of Denali potatoes showed the greatest increase (21 percent) in response to increased CO2 across all irradiance treatments, while the tuber yields of Russet and Norland were increased 18 and 9 percent, respectively. Greater plant growth from CO2 enrichment was observed under lower PPF and the shorter (12 hr) photoperiod.
Wheeler, Raymond M.↗
Low-temperature proton irradiation damage of isotropic nuclear grade IG-430 graphite
IG-430, a fine-grained, isotropic graphite grade is a promising candidate for the future Very High Temperature Reactors (VHTR). IG-430 which provides higher density, strength, and thermal conductivity, has already been developed as a graphite for next-generation HTGR, and is expected to be employed. This graphite grade, however, is lacking enough database that is needed for design. The present study aims to enhance the database with experimental data focusing on the low temperature regime (90–210 °C) by using 120–200 MeV protons to irradiate the IG-430 graphite to peak fluence of ~1.2 10 25 m –2 . It is anticipated that radiation-induced changes in the graphite properties and damage to be more pronounced in this low temperature regime than in elevated temperatures where damage annealing is taking place simultaneously. IG-430 graphite was characterized following irradiation for mechanical property changes (modulus and strength), dimensional stability and irradiation-induced growth as well as microstructural changes using high energy X-rays and different X-ray diffraction techniques. In assessing proton irradiation effects on the IG-430 graphite grade, comparison of radiation effects was made with the IG-43 grade, the un-purified version of IG-430, as well as other isotropic graphite grades. IG-430 was shown in this study to be better graphitized than other isotropic graphite grades. The study also revealed that during proton irradiation at low temperatures (~100 °C) the IG-430 exhibits stored energy release.
36 MATERIALS SCIENCE↗
Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690
Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Compositionally complex carbide ceramics: A perspective on irradiation damage
Extensive experimental and computational studies have demonstrated outstanding physical and chemical properties of the novel materials of compositionally complex carbides (CCCs), enabling their promising applications in advanced fission and fusion energy systems. This perspective provides a comprehensive overview of radiation damage behavior reported in the literature to understand the fundamental mechanisms related to the impact of multi-principal metal components on phase stability, irradiation-induced defect clusters, irradiation hardening, and thermal conductivity of compositionally complex carbides. Several future research directions are recommended to critically evaluate the feasibility of designing and developing new ceramic materials for extreme environments using the transformative “multi-principal component” concept. Compared to the existing materials for nuclear applications including stainless steels, nickel alloys, ZrC, SiC, and potentially high-entropy alloys, as well as certain other compositionally complex ceramic families. CCCs appear to be more resistant to amorphization, growth of irradiation defect clusters, and void swelling.
36 MATERIALS SCIENCE↗
Mechanistic grain growth model for fresh and irradiated UO2 nuclear fuel
Not Available
Tonks, Michael R. (ORCID:0000000213433193)↗