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At least 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 10 13 to 3 × 10 16 cm -2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluence regime, which is consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.

36 MATERIALS SCIENCE↗
DOE OSTI2017

High thermal coarsening resistance of irradiation-induced nanoprecipitates in Cu-Mo-Si alloys

Irradiation by 1.8 MeV Kr ions at room temperature (RT) was employed to induce nanoprecipitation in two Cu-Mo-Si ternary alloys, Cu 96 Mo 1 Si 3 and Cu 90 Mo 3.5 Si 6.5 . The nature and the stability of these precipitates during high temperature annealing and irradiation were studied using X-ray diffraction and transmission electron microscopy. In Cu 90 Mo 3.5 Si 6.5 , Kr irradiation at RT resulted in the formation of a high number density (>10 23 m -3 ) of Mo-Si rich nanoprecipitates, ~3 nm in size, the formation of which was attributed to Mo-Si interactions during the thermal spike phase of displacement cascades. The size of these irradiation-induced Mo-Si precipitates remained unchanged after annealing at 750 °C for 1hr. This near absence of thermal coarsening contrasted with a significant thermal coarsening of the more Mo-rich precipitates formed by direct annealing in that same ternary alloy, and with the even larger coarsening of pure Mo precipitates formed by RT irradiation in binary Cu-Mo alloys. During high temperature irradiation, precipitation is also observed, and some precipitates were significantly enriched in Mo. As for thermal annealing, those precipitates displayed increased coarsening. The remarkable thermal coarsening resistance of RT irradiation-induced Mo-Si precipitates is attributed to the suppression of Mo solubility in a Cu matrix containing Si-rich precipitates, and to the narrow precipitates size distribution generated by RT irradiation. This former point was quantified using thermodynamic calculations with the CALPHAD method.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In-situ TEM investigation of void swelling in nickel under irradiation with analysis aided by computer vision

Understanding the stability of irradiation-induced voids in materials is important for engineering material's swelling behavior under irradiation. In-situ TEM offers a spatial and temporal resolution that is suitable for investigating the evolution of voids under irradiation. However, the in-situ videos have often been too large to be analyzed manually, leaving the valuable data underutilized. For this work, we developed a deep learning-based semantic segmentation model to consistently study the growth and shrinkage of voids in nickel under 1 MeV krypton ion irradiation at various temperatures from 525 °C to 650 °C. With a foil thickness near 100 nm and ion flux of 6.3 × 10 11 ions∙ cm –2 ∙s –1 , the pre-existing voids, which were created beforehand by irradiation at 600 °C to 0.5 dpa, shrank at low temperatures and grew at high temperatures under irradiation, where the transition occurred at 575 °C (~0.5 T M ). The observed stability transition provided new insight for the shrinkage mechanism of voids under irradiation. In addition, an annealing experiment on nickel, previously irradiated at 600 °C to 3 dpa, was performed sequentially at 650 °C to 720 °C to reveal the shrinkage rate of void as a function of temperature and void size. The advantage of combining computer vision and in-situ TEM to obtain comprehensive void evolution was demonstrated.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Physiological and molecular characterization of the enhanced salt tolerance induced by low-dose gamma irradiation in Arabidopsis seedlings

Highlights: • 50-Gy gamma irradiation markedly promotes the seedling growth under salt stress in Arabidopsis. • The contents of H{sub 2}O{sub 2} and MDA are obviously reduced by low-dose gamma irradiation under salt stress. • Low-dose gamma irradiation stimulates the activities of antioxidant enzymes under salt stress. • Proline accumulation is required for the low-gamma-ray-induced salt tolerance. • Low gamma rays differentially regulate the expression of genes related to salt stress. - Abstract: It has been established that gamma rays at low doses stimulate the tolerance to salt stress in plants. However, our knowledge regarding the molecular mechanism underlying the enhanced salt tolerance remains limited. In this study, we found that 50-Gy gamma irradiation presented maximal beneficial effects on germination index and root length in response to salt stress in Arabidopsis seedlings. The contents of H{sub 2}O{sub 2} and MDA in irradiated seedlings under salt stress were significantly lower than those of controls. The activities of antioxidant enzymes and proline levels in the irradiated seedlings were markedly increased compared with the controls. Furthermore, transcriptional expression analysis of selected genes revealed that some components of salt stress signaling pathways were stimulated by low-dose gamma irradiation under salt stress. Our results suggest that gamma irradiation at low doses alleviates the salt stress probably by modulating the physiological responses as well as stimulating the stress signal transduction in Arabidopsis seedlings.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2025

Electron microscopy data on irradiation effects in glassy carbon, nuclear graphite, pyrolytic carbon, and carbon fibers

Glassy carbon, a monoatomic allotrope of carbon, is a candidate material for components in fission nuclear power systems due to its radiation tolerance. This article presents comprehensive electron microscopy data revealing the effects of neutron and electron irradiation on glassy carbon. For comparison, additional data are provided for pyrolytic graphite and carbon fibers, materials that exhibit similar structural behavior under irradiation. In situ electron irradiation experiments further illustrate the real-time microstructural evolution of glassy carbon during exposure. The dataset is organized into five parts: (1) transmission electron microscopy (TEM) micrographs of as-received and neutron-irradiated glassy carbon; (2) TEM micrographs of neutron-irradiated graphite; (3) TEM micrographs of unirradiated and irradiated carbon–carbon composites; (4) TEM micrographs of pyrolytic carbon specimens in both conditions; (5) scanning transmission electron microscopy (STEM) micrographs of as-received and neutron-irradiated glassy carbon and (6) in situ electron irradiation data of a glassy carbon particle. These datasets provide valuable insights into radiation-induced structural changes in carbon-based materials relevant to nuclear applications.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Ten-Year Survival Results of a Randomized Trial of Irradiation of Internal Mammary Nodes After Mastectomy

Purpose: To evaluate the efficacy of irradiation of internal mammary nodes (IMN) on 10-year overall survival in breast cancer patients after mastectomy. Methods and Patients: This multicenter phase 3 study enrolled patients with positive axillary nodes (pN+) or central/medial tumors with or without pN+. Other inclusion criteria were age <75 and a Karnofsky index ≥70. All patients received postoperative irradiation of the chest wall and supraclavicular nodes and were randomly assigned to receive IMN irradiation or not. Randomization was stratified by tumor location (medial/central or lateral), axillary lymph node status, and adjuvant therapy (chemotherapy vs no chemotherapy). The prescribed dose of irradiation to the target volumes was 50 Gy or equivalent. The first 5 intercostal spaces were included in the IMN target volume, and two-thirds of the dose (31.5 Gy) was given by electrons. The primary outcome was overall survival at 10 years. Disease-free survival and toxicity were secondary outcomes. Results: T total of 1334 patients were analyzed after a median follow-up of 11.3 years among the survivors. No benefit of IMN irradiation on the overall survival could be demonstrated: the 10-year overall survival was 59.3% in the IMN-nonirradiated group versus 62.6% in the IMN-irradiated group (P=.8). According to stratification factors, we defined 6 subgroups (medial/central or lateral tumor, pN0 [only for medial/central] or pN+, and chemotherapy or not). In all these subgroups, IMN irradiation did not significantly improve overall survival. Conclusions: In patients treated with 2-dimensional techniques, we failed to demonstrate a survival benefit for IMN irradiation. This study cannot rule out a moderate benefit, especially with more modern, conformal techniques applied to a higher risk population.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2013

Correlation of Clinical and Dosimetric Factors With Adverse Pulmonary Outcomes in Children After Lung Irradiation

Purpose: To identify the incidence and the risk factors for pulmonary toxicity in children treated for cancer with contemporary lung irradiation. Methods and Materials: We analyzed clinical features, radiographic findings, pulmonary function tests, and dosimetric parameters of children receiving irradiation to the lung fields over a 10-year period. Results: We identified 109 patients (75 male patients). The median age at irradiation was 13.8 years (range, 0.04-20.9 years). The median follow-up period was 3.4 years. The median prescribed radiation dose was 21 Gy (range, 0.4-64.8 Gy). Pulmonary toxic chemotherapy included bleomycin in 58.7% of patients and cyclophosphamide in 83.5%. The following pulmonary outcomes were identified and the 5-year cumulative incidence after irradiation was determined: pneumonitis, 6%; chronic cough, 10%; pneumonia, 35%; dyspnea, 11%; supplemental oxygen requirement, 2%; radiographic interstitial lung disease, 40%; and chest wall deformity, 12%. One patient died of progressive respiratory failure. Post-irradiation pulmonary function tests available from 44 patients showed evidence of obstructive lung disease (25%), restrictive disease (11%), hyperinflation (32%), and abnormal diffusion capacity (12%). Thoracic surgery, bleomycin, age, mean lung irradiation dose (MLD), maximum lung dose, prescribed dose, and dosimetric parameters between V{sub 22} (volume of lung exposed to a radiation dose ≥22 Gy) and V{sub 30} (volume of lung exposed to a radiation dose ≥30 Gy) were significant for the development of adverse pulmonary outcomes on univariate analysis. MLD, maximum lung dose, and V{sub dose} (percentage of volume of lung receiving the threshold dose or greater) were highly correlated. On multivariate analysis, MLD was the sole significant predictor of adverse pulmonary outcome (P=.01). Conclusions: Significant pulmonary dysfunction occurs in children receiving lung irradiation by contemporary techniques. MLD rather than prescribed dose should be used to perform risk stratification of patients receiving lung irradiation.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2015

Tumor Induction in Mice After Localized Single- or Fractionated-Dose Irradiation: Differences in Tumor Histotype and Genetic Susceptibility Based on Dose Scheduling

Purpose: To investigate differences in tumor histotype, incidence, latency, and strain susceptibility in mice exposed to single-dose or clinically relevant, fractioned-dose γ-ray radiation. Methods and Materials: C3Hf/Kam and C57BL/6J mice were locally irradiated to the right hindlimb with either single large doses between 10 and 70 Gy or fractionated doses totaling 40 to 80 Gy delivered at 2-Gy/d fractions, 5 d/wk, for 4 to 8 weeks. The mice were closely evaluated for tumor development in the irradiated field for 800 days after irradiation, and all tumors were characterized histologically. Results: A total of 210 tumors were induced within the radiation field in 788 mice. An overall decrease in tumor incidence was observed after fractionated irradiation (16.4%) in comparison with single-dose irradiation (36.1%). Sarcomas were the predominant postirradiation tumor observed (n=201), with carcinomas occurring less frequently (n=9). The proportion of mice developing tumors increased significantly with total dose for both single-dose and fractionated schedules, and latencies were significantly decreased in mice exposed to larger total doses. C3Hf/Kam mice were more susceptible to tumor induction than C57BL/6J mice after single-dose irradiation; however, significant differences in tumor susceptibilities after fractionated radiation were not observed. For both strains of mice, osteosarcomas and hemangiosarcomas were significantly more common after fractionated irradiation, whereas fibrosarcomas and malignant fibrous histiocytomas were significantly more common after single-dose irradiation. Conclusions: This study investigated the tumorigenic effect of acute large doses in comparison with fractionated radiation in which both the dose and delivery schedule were similar to those used in clinical radiation therapy. Differences in tumor histotype after single-dose or fractionated radiation exposures provide novel in vivo evidence for differences in tumor susceptibility among stromal cell populations.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2019

High-dose, intermediate-temperature neutron irradiation effects on silicon carbide composites with varied fiber/matrix interfaces

SiC/SiC composites are promising structural candidate materials for various nuclear applications over the wide temperature range of 300–1000 °C. Accordingly, irradiation tolerance over this wide temperature range needs to be understood to ensure the performance of these composites. In this study, neutron irradiation effects on dimensional stability and mechanical properties to high doses (11–44 dpa) at intermediate irradiation temperatures (˜600 °C) were evaluated for Hi-Nicalon Type-S or Tyranno-SA3 fiber–reinforced SiC matrix composites produced by chemical vapor infiltration. The influence of various fiber/matrix interfaces, such as a 50–120 nm thick pyrolytic carbon (PyC) monolayer interphase and 70–130 nm thick PyC with a subsequent PyC (˜20 nm)/SiC (˜100 nm) multilayer, was evaluated and compared with the previous results for a thin-layer PyC (˜20 nm)/SiC (˜100 nm) multilayer interphase. Four-point flexural tests were conducted to evaluate post-irradiation strength, and SEM and TEM were used to investigate microstructure. Regardless of the fiber type, monolayer composites showed considerable reduction of flexural properties after irradiation to 11–12 dpa at 450–500 °C; and neither type showed the deterioration identified at the same dose level at higher temperatures (>750 °C) in a previous study. After further irradiation to 44 dpa at 590–640 °C, the degradation was enhanced compared with conventional multilayer composites with a PyC thickness of ˜20 nm. Multilayer composites have shown comparatively good strength retention for irradiation to ˜40 dpa, with moderate mechanical property degradation beginning at 70–100 dpa. As a result, irradiation-induced debonding at the F/M interface was found to be the major cause of deterioration of various composites.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effects of neutron irradiation of Ti 3 SiC 2 and Ti 3 AlC 2 in the 121–1085 °C temperature range

Herein we report on the formation of defects in response to neutron irradiation of polycrystalline Ti 3 SiC 2 and Ti 3 AlC 2 samples exposed to doses of 0.14±0.01, 1.6±0.1, and 3.4±0.1 displacements per atom (dpa) at irradiation temperatures of 121±12, 735±6 and 1085±68 °C. After irradiation to 0.14 dpa at 121 °C and 735 °C, black spots are observed in both Ti 3 SiC 2 and Ti 3 AlC 2 . After irradiation to 1.6 and 3.4 dpa at 735 °C, basal dislocation loops, with a Burgers vector of b = ½ [0001] are observed in Ti 3 SiC 2 , with loop diameters of 21±6 and 30±8 nm for 1.6 dpa and 3.4 dpa, respectively. In Ti3AlC2, larger dislocation loops, 75±34 nm in diameter are observed after 3.4 dpa at 735 °C, in addition to stacking faults. Impurity particles of TiC, as well as stacking fault TiC platelets in the MAX phases, are seen to form extensive dislocation loops under all conditions. Voids are observed at grain boundaries and within stacking faults after 3.4 dpa irradiation, with extensive void formation in the TiC regions at 1085 °C. Remarkably, denuded zones on the order of 1 µm are observed in Ti 3 SiC 2 after irradiation to 3.4 dpa at 735 °C. Small grains, 3-5 µm in diameter, are damage free after irradiation at 1085 °C at this dose. The presence of the A-layer in the MAX phases is seen to provide enhanced irradiation tolerance. Based on these results, and up to 3.41 dpa, Ti 3 SiC 2 remains a promising candidate for high temperature nuclear applications.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Neutron irradiation effects on the mechanical properties of powder metallurgical processed tungsten alloys

Neutron irradiation effects on the tensile properties of Pure W, K-doped W, W–3%Re, and K-doped W–3%Re were examined under the US-Japan collaboration project PHENIX. The fission neutron irradiation experiments were carried out up to 0.74 dpa at 600 °C and 800 °C. Pure W (SR) showed irradiation hardening and loss of ductility after irradiation at 600 °C and 800 °C. K-doped W–3%Re (SR) also exhibited irradiation hardening but was ductile after irradiation. Characteristic points of the K-doped W–3%Re (SR) are small grain size and layered structure. Finally, the stress-relief treatment and the layered structure may improve the ductility of powder-metallurgy W alloys after neutron irradiation, so a combination of K-doping and Re addition would be beneficial to improve irradiation resistance.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Effects of irradiation spectrum on the microstructural and mechanical properties of bulk metallic glasses

Zr 52.5 Cu 17.9 Ni 14.6 Al 10 Ti 5 (BAM-11) and Cu60Zr 20 Hf 10 Ti 10 bulk metallic glass (BMG) specimens were ion irradiated to characterize and compare their irradiation induced microstructural and mechanical property evolution. For the ion irradiations, samples of each BMG were exposed to 9 MeV Ni 3+ and 5.5 MeV C + ions to a midrange (~1.5 μm depth) dose of 0.5 displacements per atom (dpa) at temperatures ranging from room temperature to 360 °C in order to examine potential effects associated with low (5.5 MeV C) and high (9 MeV Ni) average primary knock on energies. BAM-11 BMG samples were also irradiated by neutrons (E > 0.1 MeV) at ~70 °C to a fluence of 1.4 × 10 20 n/cm 2 (dose of 0.1 dpa). Importantly, no pronounced irradiation spectrum effects were observed for the nanoindentation mechanical properties of the two BMGs following ion irradiation. However, significant softening was observed in the BAM-11 BMG sample irradiated by neutrons, while annealing of the unirradiated samples led to a marked increase in hardening. Finally, results of the nanoindentation experiments indicate that softening is caused by irradiation induced creation of free volume defects, while the hardening is caused by their annihilation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Effects of HFIR neutron irradiation on fracture toughness properties of standard and Ni-doped F82H

F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. Here, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58 Ni, and F82H doped with 1.4% 60 Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Tensile properties of powder-metallurgical-processed tungsten alloys after neutron irradiation near recrystallization temperatures

The tensile properties of powder-metallurgical-processed Pure W, K-doped W, W-3%Re, and K-doped W-3%Re were examined after neutron irradiation up to 0.7 dpa at 910–1020 °C with a thermal neutron shield in the High Flux Isotope Reactor (HFIR). After irradiation, recrystallized Pure W (R) exhibited a brittle fracture mode, while recrystallized K-doped W-3%Re (R) exhibited a ductile fracture mode at 500 °C. K-doped W-3%Re (R) has fine grains, and hence, contains a considerable number of grain boundaries that act as sinks for irradiation defects. Solid solute Re in the W matrix could improve not only the mechanical properties of W, but also its resistance to neutron irradiation. At 500 °C, the ductility of K-doped W-3%Re after irradiation was significantly higher than that of Pure W. The irradiation at ~1000 °C did not induce hardening of stress-relieved (SR) W materials, but SR W materials tended to exhibit a decrease in the ultimate tensile strength (UTS) and an increase in total elongation (TE). Here, the softening due to the recovery and recrystallization of SR W materials and the hardening due to the formation of irradiation defect clusters were balanced during irradiation at ~1000 °C, and ductility was exhibited without an increase in strength.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Void shrinkage in 21Cr32Ni austenitic model alloy during in-situ ion irradiation

Austenitic 21Cr32Ni model alloy thin foils, previously irradiated with 5 MeV Fe ++ ions in bulk to create voids, were re-irradiated in-situ in the Intermediate Voltage Electron Microscope Facility (IVEM). The voids which had been formed under bulk-ion irradiation shrank and disappeared after in-situ Kr ion irradiation in the temperature range 50 K-713 K to an additional dose of 1 dpa. The voids were unaffected by eithersuccessive thermal annealing to 673 K and by prolonged exposure to the 200 keV electron beam at the irradiation temperature. The high void shrinkage rate observed did not change significantly for irradiation temperatures between 50 K and 713 K, suggesting that the void shrinkage process in thin foils during in-situ heavy-ion irradiation results from the interactions of displacement cascades with the voids. Finally, possible void shrinkage mechanisms under thin foil irradiation are discussed in this study.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Behavior of helium cavities in ion-irradiated W-Ni-Fe ductile-phase toughened tungsten

This study reports on the distribution of helium (He) cavities in a hot-rolled W-Ni-Fe ductile-phase toughened tungsten (DPT W) composite irradiated to a dose and a helium concentration that are comparable to those in the material after 5-year irradiation in a conceptual fusion power plant. Here, the DPT W sample consists of W particles embedded in a ductile-phase NiFeW matrix with a nominal composition of 90W-7Ni-3Fe by weight. It was hot-rolled to a thickness reduction by 87% (87R DPT W). Sequential irradiation of the material with 1.2 MeV Ni + ions to a fluence of 2.15×10 16 Ni + /cm 2 and 90 keV He + ions to 6.5×10 15 He + /cm 2 was performed at 973 K. Larger He cavities with a lower number density are observed in NiFeW than W. Helium cavities are aggregated preferentially along the NiFeW/W interphase boundary. This behavior is not observed along the W/W grain boundary under the same irradiation conditions. A similar distribution of He cavities with comparable sizes is observed in 87R DPT W irradiated with He + ions only, suggesting that the He cavity distribution is not significantly affected by pre-irradiation of Ni + ions at 973 K. The data from this study does not provide any evidence for formation of visible voids or Ni precipitates in W phase irradiated with Ni + ions only. Diffusion and clustering of He atoms and mobile vacancies during ion irradiation at 973 K are responsible for the formation and growth of the He cavities.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mechanical properties, strain hardening, and fracture behavior of ultrasonic additively manufactured Zircaloy-4 after low-temperature neutron irradiation

Ultrasonic additive manufacturing (UAM) is a solid-state, layer-by-layer advanced manufacturing process that has the potential to create custom spatially controlled composites with embedded wires and sensors for nuclear component manufacture. For this work, to assess the feasibility of using UAM for nuclear-relevant materials research, the technique was used to produce a 3.5-mm-thick Zircaloy-4 plate for irradiation testing. The UAM Zircaloy-4 specimens were irradiated in the High Flux Isotope Reactor at a target irradiation temperature of 117 °C to 2.9 displacements per atom (dpa) to assess differences in irradiation-hardening behavior as a function of alloy processing path. The UAM and reference baseplate (BP) materials increased in yield strength by 372±27 MPa and 346±21 MPa, respectively, and both suffered significant reductions in uniform and total elongation attributed to irradiation hardening at low-temperature. Although the materials had similar nanoscale defect structures, including nanoscale black dot/loop features and strain-induced dislocation channels, the UAM material’s processing-related defects resulted in accelerated strain localization and failure as demonstrated by lower post-irradiation uniform elongation of UAM specimens (0.5 %) compared to BP (1.5 %) material. The UAM material also showed considerable anisotropy in mechanical response due to crack propagation along weld boundaries, resulting in differences in strength & ductility when tested parallel and perpendicular to the prior UAM build orientation. Therefore, although the fundamental irradiation response of UAM-processed Zircaloy-4 was phenomenologically comparable to that of BP reference material, additional optimization of the UAM processing is needed to produce irradiation-resistant and nuclear-relevant materials.

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