Exploring Stability of Nanocrystalline Metals with Competing Solute Effects Under High Temperature Irradiation.
Abstract not provided.
Barr, Christopher Michael↗
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Effects of 5-Ion Beam Irradiation and Hindlimb Unloading on Metabolic Pathways in Plasma and Brain of Behaviorally Tested WAG/Rij Rats
A limitation of simulated space radiation studies is that radiation exposure is not the only environmental challenge astronauts face during missions. Therefore, we characterized behavioral and cognitive performance of male WAG/Rij rats 3 months after sham-irradiation or total body irradiation with a simplified 5-ion mixed beam exposure in the absence or presence of simulated weightlessness using hindlimb unloading (HU) alone. Six months following behavioral and cognitive testing or 9 months following sham-irradiation or total body irradiation, plasma and brain tissues (hippocampus and cortex) were processed to determine whether the behavioral and cognitive effects were associated with long-term alterations in metabolic pathways in plasma and brain. Sham HU, but not irradiated HU, rats were impaired in spatial habituation learning. Rats irradiated with 1.5 Gy showed increased depressive-like behaviors. This was seen in the absence but not presence of HU. Thus, HU has differential effects in sham-irradiated and irradiated animals and specific behavioral measures are associated with plasma levels of distinct metabolites 6 months later. The combined effects of HU and radiation on metabolic pathways in plasma and brain illustrate the complex interaction of environmental stressors and highlights the importance of assessing these interactions.
63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
Transmission electron microscopy of ion irradiated ODS MA956 samples
Oxide dispersion strengthened (ODS) alloys are promising candidate materials for the next generation advanced nuclear reactors due to their superior irradiation resistance and mechanical properties. To better understand the effect of irradiation on MA956, it is essential to study higher dose (50-100 dpa) samples, so that the general trend of microstructural evolution and the resulting radiation-hardening can be deduced. Currently, ion irradiations are considered the only way to achieve doses beyond ~50 dpa in a practical time frame relevant to alloy and welding development programs. This dataset contains TEM characterization results of ODS MA956 samples that were ion irradiated under different conditions: Sample #5 (1.25 dpa at 190℃); Sample #9 (50 dpa at 190℃); Sample #15 (1.25dpa under 320℃); and Sample 17 (25 dpa at 320℃). TEM characterization focused on the irradiation induced defects (dislocation lines and loops) using the on-zone axis bright field STEM technique. These data were collected using a FEI Tecnai G2 F30 S/TEM at Microscopy and Characterization Suite (MaCS) at Center for Advanced Energy Studies (CAES), Idaho Falls, ID. This project (ion irradiation and TEM studies) was supported by the U.S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-05ID14517 as part of Nuclear Science User Facilities award #18-14784 (PI: Ramprashad Prabhakaran, PNNL).
Prabhakaran, Ramprashad↗
Ion beam irradiation of ABO4 compounds with the fergusonite, monazite, scheelite, and zircon structures
The effects of irradiation on CaWO4, SrWO4, BaWO4, YVO4, LaVO4, YNbO4, and LaNbO4 were investigated on thin crystals using 1.0 MeV Kr ions at 50-1000 K. All of the ABO4 compounds can be amorphized with calculated damage cross sections (σa = 1/Fc0) in the range of ~0.30-1.09 × 10-14 cm2 ion-1 at zero Kelvin. The analysis of fluence-temperature data returned critical temperatures for amorphization (Tc) of 311 ± 1, 358 ± 90, 325 ± 19, 415 ± 17, 541 ± 6, 636 ± 26, and 1012 ± 1 K, respectively, for the compounds listed above. Compared with previous in situ irradiation of ABO4 orthophosphate samples using 0.8 MeV Kr ions, the Tc values of LaVO4 and YVO4 are higher than those of LaPO4 and YPO4 by 82 K and 124 K, respectively. The Tc values of the three scheelite structures, CaWO4, SrWO4, and BaWO4, indicate that they are the most radiation tolerant compounds under these conditions. The A-B cation anti-site energies, EfAB, determined by DFT range from 2.48 to 10.58 eV and are highly correlated with the A-B cation ionic radius ratio, rA/rB, but are not correlated with Tc across the different structure types, suggesting that the formation and migration energies of Frenkel defects play a more important role in damage recovery in these compounds. We also discuss the role of cation and anion charge/iconicity as determined by DFT. ABO4 compounds with the zircon structure and B = P or V have a distinct advantage over those with B = Si as the damaged regions do not appear to be significantly affected by polymerization of (PO4) 3- or (VO4) 3- groups which might stabilize the amorphous fraction and ultimately lead to phase separation as observed in zircon (ZrSiO4).
amorphization↗
A quantitative method to determine the region not influenced by injected interstitial and surface effects during void swelling in ion-irradiated metals
We propose and demonstrate a microstructurally-based experimental method to quantitively determine the depth regions in self-ion-irradiated metals that are affected by the injected interstitial effect and various surface effects, focusing on the choice of safe analysis zones to minimize the impact of these phenomena. The goal is to define the depth ranges where extracted data can be confidently applied to ion-neutron correlations for reactor application. Since ion energies in the range of 1–5 MeV are most frequently employed by the radiation effects community, irradiations were conducted at four energies in this range, all proceeding at 475 °C. The experiment was conducted on relatively pure single crystal iron to focus only on physical phenomena, avoiding the influence of possibly confounding chemical or segregation processes. Care was also taken to minimize the influence of other physical factors such as crystalline orientation. It was shown that, at 475 °C, ion energies of ≤1 MeV were too shallow in penetration and did not yield a safe depth range, but irradiations at 2.5 MeV and above yielded useful safe zones with predicted swelling behavior becoming independent of ion energy. The surface-affected zone width was found to be roughly twice that of the void-denuded zone width and to be independent of accumulated displacement dose. The largest injected-interstitial effect arises from the injected interstitial depression of void nucleation and growth. The interstitial-affected region starts at about one half of the projected range and does not show any “spreading” of its influence in depth as the peak damage level increases from 50 to 100 dpa. This study provides some confidence that enhances the credibility of ion simulation when applied to prediction of void swelling in neutron environments.
36 MATERIALS SCIENCE↗
Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten
The effect of mixed spectra and thermal neutron shielded irradiation on tungsten was evaluated with plasma exposure in the tritium plasma experiment followed by thermal desorption spectroscopy, X-ray photoelectron spectroscopy, and transmission electron microscopy. The two different irradiation campaigns were performed at the High Flux Isotope Reactor to 0.39–0.74 displacement per atom (dpa) in the 894–1379 K temperature range. A neutron spectrum influence on the void size and void number density was not observed. However, a strong correlation was found between void size and void number density with temperature, but not with dpa in the limited dpa range of this study. Thermal neutron shielding significantly reduced the transmutation to Re+Os. Higher irradiation temperature will lead to larger voids with lower number density, which reduces deuterium retention. In conclusion, grain growth was also observed for high-temperature irradiation of over ~1300 K within the limited grains visible in the transmission electron microscopy specimens.
36 MATERIALS SCIENCE↗
Ultra-high-dose-rate FLASH and Conventional-Dose-Rate Irradiation Differentially Affect Human Acute Lymphoblastic Leukemia and Normal Hematopoiesis
Ultra-high-dose-rate FLASH radiation therapy has been shown to minimize side effects of irradiation in various organs while keeping antitumor efficacy. This property, called the FLASH effect, has caused enthusiasm in the radiation oncology community because it opens opportunities for safe dose escalation and improved radiation therapy outcome. Here, we investigated the impact of ultra-high-dose-rate FLASH versus conventional-dose-rate (CONV) total body irradiation (TBI) on humanized models of T-cell acute lymphoblastic leukemia (T-ALL) and normal human hematopoiesis.
62 RADIOLOGY AND NUCLEAR MEDICINE↗
In-situ ion irradiation study of alloy 709 stainless steels with different processing histories
Alloy 709 stainless steel (A709 SS) has gained an increased interest in nuclear applications due to its advanced properties at high temperatures over the widely used Type 316 SS. However, how this alloy behaves under irradiation with concurrent environmental factors, such as mechanical deformation and/or thermal annealing, has not been studied. In this work, we used in-situ ion irradiation inside a transmission electron microscope (TEM) to investigate the effects of irradiation on the microstructures of A709 SS samples with different prior processing histories, including as-received, annealed, and creep-deformed. It was found that, compared to room temperature irradiation, 600 degrees C irradiation led to larger dislocation loop sizes and lower loop densities. Furthermore, with 600 degrees C irradiation, the creep-deformed sample, which had the highest precipitate density and the highest dislocation density, had the smallest loop size and the lowest loop density compared to others. This observation has been rationalized with quantitative evaluation of the sink strength factors in each sample. This study shows that in-service deformation can significantly affect the irradiation performance in structural materials at reactor operating conditions.
36 MATERIALS SCIENCE↗
Comparison of PM-HIP to forged SA508 pressure vessel steel under high-dose neutron irradiation
Not Available
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Correlated transmission electron microscopy and atom probe tomography characterization of ion irradiated Ni-based alloy Hastelloy N
Ion irradiation of Hastelloy N was conducted to better characterize the effects of irradiation on Hastelloy N using modern tools compared to studies done in the 1950s. The 2 MeV Ni + ion irradiation at 600 °C of Hastelloy N has been investigated using atom probe tomography, transmission electron microscopy and energy dispersive spectroscopy. Irradiation is found to promote formation of nanoscale M 2 C carbides over the thermodynamically favored M 6 C. Segregation of Si to dislocation loops and grain boundaries was also evident and may have assisted in formation of M 2 C. These microstructural changes result in a 20 % hardness increase caused by irradiation alone. These observations are useful in the design of new materials better suited for the harsh environment of a molten salt reactor.
Atom probe tomography↗
α Irradiation Response on the Electronic Transport Properties of p -B 12 P 2
B 12 P 2 , a wide bandgap semiconductor, has been previously reported to self-heal from high-energy electron bombardment as determined from transmission electron microscopy. However, the effect of irradiation on the electronic transport properties of B 12 P 2 has not been studied. We report electrical transport measurements on B 12 P 2 after α irradiation. The temperature-dependent Hall effect was measured on p-type B 12 P 2 after successive α irradiation from 210 Po over a fluence range of 1 × 10 11 –1 × 10 13 α/cm 2 to test for self-healing of B 12 P 2 to α bombardment. Prior to irradiation, B 12 P 2 displayed electrical transport characteristics of both hopping and impurity/defect band conduction due to background impurities and growth-induced defects. Fluences of 5 × 10 12 α/cm 2 or greater caused a decrease in the hole mobility and an increase in defect band conduction, suggesting defect accumulation in B 12 P 2 .
36 MATERIALS SCIENCE↗
Investigation of the Anisotropy of Y-Ba-Cu-O Thin Films With Photoemission Electron Microscopy
Understanding the anisotropy in the electrical properties of YBa2Cu3O7-δ (YBCO) is important for understanding the fundamental mechanism of superconductivity in unconventional superconductors. Additionally it is also essential for achieving uniform and controllable critical currents and resistances in YBCO Josephson junctions. Here, this study correlates work function measurements of different intrinsic regions on YBCO thin films, obtained through PhotoEmission Electron Microscopy (PEEM), with in plane tunneling data and helium ion imaging. YBCO thin films, 40 nm in thickness, were selectively irradiated with 300 keV helium using a Pelletron to investigate the effects of irradiation compared to unirradiated regions. Post-irradiation PEEM images revealed distinct contrasts between irradiated and non-irradiated regions, suggesting a metal-to-insulator transition. In non-irradiated areas, a domain pattern was detected, showing a 25 meV difference in work function, attributed to CuO and BaO rich surface regions. A similar domain structure was observed using high-resolution helium ion imaging; however, the helium ion beam can damage the YBCO, making PEEM a valuable non-destructive alternative. This work highlights the correlation between PEEM measurements and in-plane tunneling properties of YBCO, providing insights into material orientation and its impact on Josephson junction performance. Furthermore, it offers a non-destructive approach to understanding the anisotropy of cuprate materials, its superconducting properties, and the factors affecting Josephson junction performance.
Qu, Rochelle Y. [Univ. of California, Riverside, C↗
Radial Deconsolidation and Leach-Burn-Leach of Eight As-Irradiated AGR-3/4 TRISO Fuel Compacts
Eight as-irradiated AGR-3/4 fuel compacts were subjected to destructive post-irradiation examination via radial-deconsolidation-leach-burn-leach (RDLBL) at INL. The RDLBL process deconsolidated the compacts in multiple, radial steps, followed by a final, single-step axial deconsolidation. The samples generated at each step were analyzed for isotopes of key fission products and actinides. After each deconsolidation step, the compact volume was assessed, and this was used to normalize the measured quantity of nuclides of interest to give a volumetric concentration as a function of radial position within the compact. The total inventories of measured fission products and actinides and the radial concentration profiles were compared among the eight compacts deconsolidated at INL and four other as-irradiated compacts examined at ORNL. The results were analyzed for the effects of irradiation temperature. These results will be used for comparisons with fission product transport models and as input from which fission product diffusivities can be calculated.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Assessment of MiniFuel Subcapsule Design Recommendations on Previous Experiments
MiniFuel describes the class of separate effects nuclear fuels irradiation experiments that have been conducted in the High Flux Isotope Reactor (HFIR) since 2018. These experiments comprise a stack of six fuel-bearing subcapsules contained in a stainless steel target housing that is in contact with HFIR coolant on its exterior. All MiniFuel targets have a near-standardized architecture, and the primary design variables that change between experiments are the radial gap size between the subcapsule and housing and the target fill gas composition. Finite element heat transfer models are used to determine the optimum gas composition and gap sizes, and recent studies were performed to identify model parameters that contribute the most uncertainty to fuel specimen temperature predictions. That work, which is referenced herein, also recommended a set of design modifications to the subcapsule internal architecture and assembly process. These modifications are intended to reduce fuel temperature uncertainty in future experiments. In this report, the subcapsule design modifications were retroactively applied to a previously conducted MiniFuel experiment to determine how these changes affect the established safety and performance envelope of the experimental capability. These effects were determined in two steps. First, the modifications were applied to the subcapsule design without any other changes to determine their isolated effect on the predicted fuel specimen temperatures. This portion of the analysis showed that fuel temperatures were modestly reduced because the implemented changes improved heat transfer efficacy. Next, traditional MiniFuel design activities (i.e., sizing the gas gaps and determining the fill gas composition) were reperformed, and they confirmed that the original desired fuel temperatures could be achieved while remaining within established safety limits. Therefore, this report demonstrates improved performance resulting from the subcapsule modifications, which mitigate uncertainty while meeting the objectives of past experiments. An additional benefit of the design changes is reduced sensitivity of the fuel temperature to the evolving flux spectrum in HFIR, leading to more stable temperatures and enhanced utility of MiniFuel as a separate effects irradiation platform.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Microstructural Characterization of Ion Irradiated ODS MA956 Alloy
Oxide Dispersion Strengthened (ODS) alloys are considered as promising nuclear materials for the next generation advanced nuclear reactors due to their improved elevated temperature mechanical properties and radiation damage resistance. MA956 is a commercial ODS alloy with superior oxidation resistance and creep strength at elevated temperatures. The improved properties for ODS alloys can be attributed to the presence of dispersed nanosized oxide particles. These oxide particles can act as barriers to dislocation motion and provide additional sinks and sites for point defect recombination. Hence, their stability can greatly affect the properties of the irradiated ODS alloys. Besides, irradiation will induce defects into the ODS alloys, and these defects can develop further as irradiation dose increases. Furthermore, the defects propagation can significantly undermine the mechanical properties of the bulk materials, resulting in irradiation hardening and lower ductility. Therefore, it is necessary to have a good understanding of the defect’s evolution with irradiation. Hence, efforts are ongoing to understand the effect of ion irradiation on ODS MA956 with emphasis on oxide particles and evolution of defects as a function of dose (2.5, 50 and 100 dpa).
07 ISOTOPE AND RADIATION SOURCES↗
Plasticity in irradiated FeCrAl nanopillars investigated using discrete dislocation dynamics
Here, in this paper, we investigate plasticity in irradiated FeCrAl nanopillars using discrete dislocation dynamics simulations (DDD), with comparisons to transmission electron microscopic (TEM) in situ tensile tests of ion- and neutron-irradiated commercial C35M FeCrAl alloy. The effects of irradiation-induced defects, such as a/2<111> and a<100> type loops and composition fluctuations representative of phase separation in irradiated FeCrAl alloys, are investigated separately as well as superposed together in simulations. We explore the effects of defects on the stress-strain behavior, specifically yield strength and hardening response, of FeCrAl nanopillars. Our simulations confirm the widely accepted fact that irradiated alloys exhibit a stress-strain response with higher yield strength as compared to unirradiated alloys. However, our DDD calculations reveal an atypical superposition of the hardening contributions due to composition inhomogeneity and irradiation loops wherein composition inhomogeneity annihilates the hardening due to irradiation loops at small scales. As a result, we observe that the yield strength in irradiated alloys, after taking into consideration the effects of both composition inhomogeneity and irradiation loops, is smaller than the yield strength of the alloys with only irradiation loops and is approximately same for the alloy with composition inhomogeneity alone. This is referred to as “destructive interference” between the hardening contributions due to composition fluctuations and irradiation loops in the paper. We also identify this destructive interference in the superposition in our parallel TEM in situ tensile tests on unirradiated, ion-irradiated, and neutron-irradiated C35M FeCrAl alloy. This destructive interference in the hardening contributions contrasts with the dispersed barrier hardening (DBH) models widely utilized by the experimental community to model the hardening contributions due to different irradiation induced defects. The effects of the loading orientations on the yield strength and hardening are investigated and the mechanisms for the hardening in irradiated FeCrAl alloys are also reported.
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↗