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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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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Understanding of corrosion mechanisms after irradiation: effect of ion irradiation of the oxide layers on the corrosion rate of M5 - 2016-0081

Irradiation damage in fuel cladding material is mainly caused by the neutron flux that results from fission reactions occurring in the fuel. To avoid the constraints inherent in handling radioactive material, the irradiation effects on the corrosion resistance of zirconium alloys can be studied by irradiating the materials with ions. We performed an original experiment using ion irradiation to more specifically study the influence of irradiation damage in the oxide on the corrosion rate of M5{sup R}. It has been established that irradiation with a 1.3-MeV helium ion at a fluence of 10{sup 17} cm{sup -2} results in significant modifications of oxide properties, oxygen diffusion flux, and oxidation kinetics, as evidenced by Raman spectroscopy, secondary ion mass spectrometry (SIMS) analyses, and measurements of mass gains. A newly identified Raman vibration band at 712 cm{sup -1} was linked to the presence of irradiation defects and allowed the evolution of their concentrations to be followed. The oxygen diffusion flux was significantly reduced after irradiation partly due to a surface concentration decrease of oxygen. The defects remained present in the oxide after 100 days of annealing in pressurized water reactor (PWR) conditions and were thus very stable in PWR conditions, which probably means that these defects would be stable in the reactor. According to the kinetics and in agreement with the results obtained by SIMS analyses, the oxidation rate was significantly reduced after ion irradiation, and this effect remained beyond 100 days in agreement with the high stability of irradiation defects in PWR conditions. An original model described quite well the oxidation kinetic results. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2017

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

Here, complex microstructure changes occur in nuclear fuel and structural materials due to the extreme environments of intense irradiation and high temperature. This paper evaluates the role of the phase field method in predicting the microstructure evolution of irradiated nuclear materials and the impact on their mechanical, thermal, and magnetic properties. The paper starts with an overview of the important physical mechanisms of defect evolution and the significant gaps in simulating microstructure evolution in irradiated nuclear materials. Then, the phase field method is introduced as a powerful and predictive tool and its applications to microstructure and property evolution in irradiated nuclear materials are reviewed. The review shows that (1) Phase field models can correctly describe important phenomena such as spatial-dependent generation, migration, and recombination of defects, radiation-induced dissolution, the Soret effect, strong interfacial energy anisotropy, and elastic interaction; (2) The phase field method can qualitatively and quantitatively simulate two-dimensional and three-dimensional microstructure evolution, including radiation-induced segregation, second phase nucleation, void migration, void and gas bubble superlattice formation, interstitial loop evolution, hydrate formation, and grain growth, and (3) The Phase field method correctly predicts the relationships between microstructures and properties. The final section is dedicated to a discussion of the strengths and limitations of the phase field method, as applied to irradiation effects in nuclear materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Decoupling and tuning competing effects of different types of defects on flux creep in irradiated YBa 2 Cu 3 O 7-δ coated conductors

YBa 2 Cu 3 O 7-δ coated conductors (CCs) have achieved high critical current densities (J c ) that can be further increased through the introduction of additional defects using particle irradiation. However, these gains are accompanied by increases in the flux creep rate, a manifestation of competition between the different types of defects. In this paper, we study this competition to better understand how to design pinning landscapes that simultaneously increase J c and reduce creep. CCs grown by metal organic deposition show non-monotonic changes in the temperature-dependent creep rate, S(T). Notably, in low fields, there is a conspicuous dip to low S as the temperature (T) increases from ~20 to ~65 K. Oxygen-, proton-, and Au-irradiation substantially increase S in this temperature range. Focusing on an oxygen-irradiated CC, we investigate the contribution of different types of irradiation-induced defects to the flux creep rate. Specifically, we study S(T) as we tune the relative density of point defects to larger defects by annealing both an as-grown and an irradiated CC in O 2 at temperatures T A = 250 °C–600 °C. Finally, we observe a steady decrease in S(T > 20 K) with increasing T A , unveiling the role of pre-existing nanoparticle precipitates in creating the dip in S(T) and point defects and clusters in increasing S at intermediate temperatures.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2018

Raman spectroscopy of neutron irradiated silicon carbide: Correlation among Raman spectra, swelling, and irradiation temperature

The effects of neutron irradiation on microstructural evolution and the resultant changes in physical and mechanical properties are of critical importance for the development of silicon carbide (SiC) materials for nuclear applications. This study neutron–irradiated β–SiC under a wide range of conditions at temperatures between 235 and 750°C and neutron doses of 0.01–11.8 displacements per atom, and then evaluated the effects on the SiC structure using Raman spectroscopy. The SiC optical phonon lines were shifted to lower wavenumbers by irradiation. Correlations were found among the wavenumber of the longitudinal optical phonon line, irradiation–induced swelling, and irradiation temperature. The peak shift also correlated indirectly with decreasing thermal conductivity of irradiated SiC. The irradiation–induced peak shift is explained by combinations of lattice strain, reduction of the elastic modulus, and other factors including decreasing coherent domain size. Lastly, these findings bridge irradiation–induced microstructural changes and property changes and illustrate how Raman spectroscopy is a useful tool for nondestructively assessing irradiated SiC materials for nuclear applications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Post Irradiation Examination Results of Irradiated Yttrium Hydride

Department of Energy’s (DOE’s) Microreactor program (MRP) aims to provide the fundamental data to enable the development of microreactors. As such, material property data of critical materials for microreactor technologies are researched. Because substoichiometric yttrium dihydride (YHx, where x<2) is considered as a potential solid neutron moderator, its material property data has been combined in the Advanced Moderator Material Handbook which includes thermodynamic and thermophysical properties of YHx with the exception of irradiated material’s properties due to limited PIE. To fill the knowledge gap for the irradiated YHx, specimens and irradiation capsules were prepared at Los Alamos National Laboratory (LANL). Specimens were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory’s (INL’s). post-irradiation examination (PIE) was performed at INL’s Materials and Fuels Complex (MFC). This report compiles the PIE results of irradiated YHx specimens through fiscal years 2022 and 2023 (FY22-23) . The PIE data will directly be incorporated into the newer version of the Advanced Moderator Material Handbook. The main takeaways include that (i) the geometrical stability and mechanical integrity of YHx was intact with couple exceptions after high-temperature irradiations (600-800°C), (ii) hydrogen content variation due to manufacturing or irradiation in YHx caused visible surface discoloration, that is also related to the microstructural changes, (iii) qualitative comparisons of PIE methods implied that H retention was significantly higher at 600°C as compared to 800°C, (iv) thermal properties included signatures correlated with the H loss or re-gain, (v) importance of manufacturing readiness and initial as-manufactured specimens history was emphasized, (vi) the needs of targeted irradiations focusing on temperature and time parameters and very targeted PIE were specified.

08 HYDROGEN↗
DOE OSTI2021

Evolution of {delta}ferrite in a CF3 cast stainless steel upon neutron irradiation to 3, 5, 10, 20, and 40 dpa

The microstructural evolution of delta ferrite in a CF3 cast stainless steel irradiated to 3, 5, 10, 20, and 40 dpa with two dose rates was studied with atom probe tomography (APT) and transmission electron mi-croscopy. Spinodal decomposition and G-phase precipitates induced by neutron irradiation and thermal aging were quantified systematically. The neutron irradiation significantly enhances the spinodal decom-position in delta ferrite as both the spinodal wavelength (i.e. a characteristic repeat distance) and am-plitude (i.e. magnitude of elemental concentration fluctuation) increase after irradiation. The wavelength only varies slightly when the dose increases from 5 to 40 dpa, while the amplitude increases dramatically from 10 to 20 dpa and starts to saturate before 20 dpa. The dose rate and irradiation temperature also have a notable effect on the spinodal wavelength and amplitude. The study shows that higher dose rate promotes a larger wavelength at given irradiation doses. Regarding the G-phase precipitates, a slightly lower irradiation temperature would result in a smaller mean size of G-phase precipitates, while a lower irradiation dose rate would lead to a larger G-phase precipitates at a given temperature and dose. Over-all, the spinodal decomposition and G-phase precipitates in the delta ferrite continue to evolve with the increasing dose beyond 10 dpa. This study confirms that the formation of G-phase precipitates at the in-terdomain region between alpha and alpha-prime is facilitated by the Si and Mo atoms rejected from Fe rich alpha phase and the Ni and Mn atoms rejected from Cr rich alpha-prime phase. Neutron irradiation plays a dominant role in the ferrite instability, and the effect of prior thermal aging at 400 degrees C for 10,000 hours is negligible as the dose is 3 dpa and above.

duplex stainless steel↗
DOE OSTI2021

Irradiation hardening and ductility loss of Eurofer97 steel variants after neutron irradiation to ITER-TBM relevant conditions

Ten Eurofer97 steel variants, produced by non-standard fabrication-processing routes and modified alloying chemistries, were studied by neutron irradiations in the high flux isotope reactor. The irradiations were performed to ITER-TBM relevant conditions of ~255–350 °C, 2.94–3.24 dpa. We quantified the irradiation-induced degradation of the steels using mechanical property tests. All the steels suffered from irradiation hardening, where a significant increase in Vickers microhardness and yield stress (σ YS ) occurred, accompanied with severe loss of tensile elongation. The extent of hardening was material dependent. For T irr = 300±30 °C, most steels showed σ YS increase in the range of ~30% to as high as ~66%, except for a low temperature tempered steel with σYS increase below 15%. Despite large losses in elongation, most failures were ductile. Significant post-necking ductility was retained with reduction in area (RA) between 65–75%, but <50% for low temperature tempered steels. The ultimate tensile stress to yield stress (σ UTS /σ YS ) ratios decreased significantly after irradiation, highlighting irradiation-induced strain hardening capacity reduction. No major effect of irradiation on the plastic instability stress (σ PIS ) and true fracture stress of the steels was observed. By comparing the tensile stresses in true stress units and with literature, the results suggest that RAFM steel designing should target materials with a large separation between σ PIS and σ YS , to ensure the materials can maintain large work hardening and uniform deformation capability after irradiation. The tensile data of the steels additionally revealed a compelling evidence of an inverse trend between the change in RA and increase in σ YS of the neutron irradiated Eurofer97 type steels.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Microstructural evolution of 3C-SiC exposed to simultaneous neutron irradiation and helium implantation

The synergism of transmutant helium and irradiation defects is critically important to understand the performance of SiC when in service in nuclear environments. In this study, we exposed 3C-SiC in contact with a 2 μm Ni foil to neutron irradiation at 500 °C to 29 dpa in the high flux isotope reactor to simulate the simultaneous introduction of helium and irradiation defects into SiC. The combination of TEM observations and thermal desorption measurements helps to elucidate the impact of helium on the microstructural evolution in 3C-SiC. Thermal helium desorption spectra from ion-implanted and neutron-irradiated SiC showed a completely different desorption behavior in terms of the peak position and helium desorption flux. The identification of possible helium trapping sites was attempted by applying the first order dissociation model to the measured thermal helium desorption spectra. TEM observations of the neutron-irradiated samples showed that the presence of helium facilitated the stabilization of the defect clusters and promoted the formation of visible helium bubbles in SiC subject to simultaneous neutron irradiation and in-situ He implantation. Following the subsequent heat treatment during the thermal desorption measurements, large faceted helium bubbles were found along grain boundaries in the neutron-irradiated SiC with in-situ He implantation, while the helium bubbles in the grain interior were relatively smaller than those found in the neutron-irradiated sample without in-situ He implantation due to the reduced mobility of He-defect clusters. In conclusion, the He contents contained in the neutron-irradiated samples were quantified based on the experimental data from TEM and thermal desorption measurements.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Irradiation of UC1+x kernels using the MiniFuel vehicle: Microstructure, phase analysis, and initial post-irradiation examination

Uranium carbide is a candidate fuel form for a wide range of advanced reactors, including larger Generation IV reactors as well as small modular reactors and microreactors. However, its commercial deployment timeline faces challenges via traditional qualification approaches. To address this issue, an accelerated fission rate irradiation test was performed to collect basic fuel performance data to inform fuel performance models and potential future integral tests. Hyperstoichiometric UC (UC1+x) kernels were irradiated in the High Flux Isotope Reactor using the MiniFuel irradiation vehicle. The test matrix spanned two temperature regimes (700 °C and 800 °C) and burnup levels (1.8 % FIMA and 2.8 % FIMA). Between 21 and 63 kernels were tested at each unique temperature and burnup condition. As-fabricated microstructural analysis revealed a multiphase composition with UC, UC2, UC2−y, and U-C-O bearing phases for the irradiated kernels. Following irradiation, fission gas release, swelling, and microstructures were analyzed. Measured fission gas release was below 5 % for all irradiation conditions, reaching a maximum at the highest temperature and burnup condition. A binary swelling response was observed; the lower burnup and temperature conditions resulted in negligible swelling, but the higher burnup and temperature conditions produced significant anisotropic swelling and densification in a subset of kernels. The basic microstructural exams of kernels following irradiation were not capable of showing a correlation between kernels that exhibited excessive swelling and those that did not. Characterization of a subset of samples using the Advanced Photon Source and more detailed microstructural examination of unirradiated kernels revealed that a subset of kernels contained very high UC2 phase fractions. The anomalous swelling response is hypothesized to have been driven by this chemical variation. The results of this irradiation highlight the potential of accelerated fission rate irradiation testing to explore such behaviors and inform the development of fuel specifications.

Adorno Lopes, Denise [ORNL] (ORCID:000900023705987↗
DOE OSTI2019

Automated Quantitative Analysis of Extended Irradiation Defects - Dislocations, Voids and Precipitates in Neutron Irradiated HT-9 Steel

Structural alloys that make up the core components of nuclear reactors are exposed to radiation that can substantially alter the microstructure and mechanical properties of the material. Irradiation can cause the formation of extended defects including dislocation line and loops, voids and various types of precipitates. To understand the irradiation effects, it is critical to characterize these defects individually and collectively to quantify how they evolve throughout irradiation stages. Transmission electron microscopy (TEM) is a standard tool for providing extensive size, number density and spatial distribution of extended irradiation defects, however, manual measurement of defect size and number density requires time-intensive effort to analyze enough defects that provide a statistically meaningful result and are prone to errors of subjectivity. In this work, we developed a set of MATLAB algorithms offering robust and automated identification and quantification of extended irradiation defects like dislocation lines, voids and precipitates commonly observed in neutron irradiated cladding alloys [1]. Using a neutron-irradiated HT-9 martensitic alloy irradiated at 412°C to 111.8 dpa, we demonstrate the capability of these algorithms for quantitative defect analysis. Figure 1 presents defect images of the HT-9 using an advanced diffraction contrast imaging scanning transmission electron microscopy (DCI STEM) that suppresses bend contours and provides better defect contrast compared to conventional TEM [2]. The first step of defect analysis is defect identification, a process that can be digitally achieved by contrast thresholding (for simple defect contrast) and by machine learning computer vision for defect semantic segmentation [3]. Here, labeled binary defect images (Figures 2a and 2c) were used. For dislocation density quantification, we employed an optimized circular-grid intersection method first introduced by Dr. John Steeds [4]. This method is shown in Figures 2a and 2b, where the number of intersections of dislocation lines with circular traces was measured. This produces an unbiased sampling of dislocation lines, regardless of their orientation. This process was optimized with automated digital intersection number and grid length registration. Interestingly, it was found that a relatively large number of circles in a set circular gird is was needed to produce an invariant result (Figure 2b), with a low systematic measurement error. This demonstrates the main advantage of computer-based defect analysis, in performing repetitive operations quickly with a reduced systematic error. For the quantitative analysis of voids and precipitates a challenge encountered with automated defect analysis is feature overlap, particularly when the density of voids/precipitates is high. To mitigate this problem, as shown in Figure 2d, we modified the popular circular Hough transformation and treated ‘circular shape’ as the particle descriptor for voids/precipitates, efficiently resolving the overlapping particles. In summary, this automated digital quantification together with machine learning enabled defect identification [3] opens new opportunities for standardizing the characterization workflow for quantitative analysis of extended irradiation defects.

Sainju, Rajat↗
DOE OSTI2024

Understanding Formation of Irradiation-Induced Defects through 4D-STEM, Electron Tomography, and WBDF-STEM

A major challenge in advancing nuclear materials for next-generation fission and proposed fusion reactors is to comprehensively understand the formation of irradiation-induced defects. Here it is essential to correlate the evolution of irradiation-induced defects and the degradation of mechanical properties, as they collectively dictate the material's lifespan and ensure nuclear safety. Scanning transmission electron microscopy (STEM) based techniques have emerged as indispensable tools for irradiation-induced defect characterization, offering high spatial resolution imaging and chemical analysis, such as electron energy loss spectroscopy (EELS) and energy dispersive X-ray spectroscopy (EDXS). These techniques have been effectively used to obtain an atomic-scale view of the defect structure. Recent advances in electron microscopy, particularly in 4D-STEM, offer detailed insight into microstructural evolution by capturing full 2D diffraction patterns at every pixel position. Using high-speed direct electron detectors, this technology generates a four-dimensional dataset, overcoming the limitations of traditional STEM imaging.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Final Report on MEGAPIE Target Irradiation and Post-Irradiation Examination

Megawatt pilot experiment (MEGAPIE) was successfully performed in 2006. One of the important goals of MEGAPIE is to understand the behaviour of structural materials of the target components exposed to high fluxes of high-energy protons and spallation neutrons in flowing LBE (liquid lead-bismuth eutectic) environment by conducting post-irradiation examination (PIE). The PIE includes four major parts: non-destructive test, radiochemical analysis of production and distribution of radionuclides produced by spallation reaction in LBE, analysis of LBE corrosion effects on structural materials, T91 and SS 316L steels, and mechanical testing of the T91 and SS 316L steels irradiated in the lower part of the target. The non-destructive test (NDT) including visual inspection and ultrasonic measurement was performed in the proton beam window area of the T91 calotte of the LBE container, the most intensively irradiated part of the MEGAPIE target. The visual inspection showed no visible failure and the ultrasonic measurement demonstrated no detectable change in thickness in the beam window area. Gamma mapping was also performed in the proton beam window area of the AlMg 3 safety-container. The gamma mapping results were used to evaluate the accumulated proton fluence distribution profile, the input data for determining irradiation parameters. Radiochemical analysis of radionuclides produced by spallation reaction in LBE is to improve the understanding of the production and distribution of radionuclides in the target. The results demonstrate that the radionuclides of noble metals, 207 Bi, 194 Hg/Au are rather homogeneously distributed within the target, while radionuclides of electropositive elements are found to be deposited on the steel-LBE interface. The corrosion effect of LBE on the structural components under intensive irradiation was investigated by metallography. The results show that no evident corrosion damages. However, unexpected deep cracks were found in the EBW (electron beam weld) of the LBE container in the intensive irradiation zone of the target, which should be formed during irradiation. In the SS 316L steel of the flow guide tube, inclusions or precipitates enriched with O, Si, S, Ca, Ti and Mn were observed. Many of them are very long, up to a few mm, and located on grain boundaries along the extrusion direction of the tube. The degradation of the mechanical properties of the T91 and SS 316L steels has been investigated by conducting tensile tests on the specimens extracted from the T91 and SS 316L components in the intensive irradiation region. The results obtained from the proton beam window of the T91 calotte exhibit a good ductility of T91 steel after irradiation at 6-7 dpa (displacement per atom) in contact with flowing LBE.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Interpretation of Ion Irradiation and Neutron Irradiation Damage in Additively Manufactured 316 Stainless Steel using Multiscale Modeling

The accelerated adoption of nuclear energy necessitates advanced manufacturing technologies, such as additive manufacturing, to meet heightened supply chain requirements and support innovative reactor technologies. Due to the unique microstructural characteristics of additively manufactured materials under distinct solidification conditions, comprehensive evaluation of their performance in reactor environments is essential. The Advanced Materials and Manufacturing Technologies program under the Department of Energy's Office of Nuclear Energy focuses on understanding the irradiation performance and damage evolution of laser powder bed fusion 316 stainless steel, with an emphasis on integrating ion and neutron irradiation data to accelerate the development and qualification of materials for advanced nuclear reactor applications. While ion irradiation is a cost- and time-effective method, modeling and simulation are required to interpret the data for the broader range of irradiation conditions encountered in advanced reactors. In fiscal year 2025, integrated multiscale modeling and simulations were conducted to assess irradiation damage in additively manufactured 316 stainless steel. Key outcomes include predictions of chromium enrichment at grain boundaries, nickel enrichment at dislocation cell walls and void surfaces, and heterogeneous void evolution under ion and neutron irradiation conditions. Cluster dynamics simulations revealed the coarsening of voids at high irradiation temperatures and the suppression of void growth by high network dislocation density, while also demonstrating significant growth and coarsening of voids and self-interstitial atom loops at low dose rates. Machine learning-accelerated atomistic simulations highlighted the impact of the local environment and chromium concentration on vacancy diffusivity, providing key insights on the influence of composition on void swelling and radiation-induced segregation. Additionally, molecular dynamics simulations demonstrated the presence of defect production bias and a significant effect of carbon content on defect cluster behavior. These combined efforts aim to predict the performance of additively manufactured materials under various reactor conditions, supporting their qualification for nuclear reactor applications by interpreting ion irradiation data. This report underscores the potential of integrated multiscale modeling to analyze ion irradiation data in the effort to accelerate the qualification of additively manufactured materials for nuclear reactor components.

316 stainless steel↗
DOE OSTI2021

Damage development in neutron-irradiated concrete in a test reactor: Hygro-thermal and mechanical simulations

This paper reports the development of a 3D mesoscale hygro-thermal-mechanical simulation approach to predict damage in concrete irradiated in a test reactor. This framework, developed in MOOSE, considers the effects of elevated temperature, moisture content, and high neutron fluence (energy threshold, E > 0.1 MeV) on the mortar and aggregates separately. The first-stage simulation implements hygro-thermal analysis to determine the temperature and RH inside the specimen as a function of imposed radiation energy. These are used as inputs to the second stage, which considers radiation-induced volumetric expansion (RIVE) of aggregates, and creep, shrinkage, and stress-strain response of mortar to predict the expansion, stresses, and damage in specimens made using different coarse aggregates and subjected to different irradiation times. The irradiation time-dependent damage in the mortar is expressed using an isotropic damage parameter. This multi-physics model serves as a predictive tool for damage quantification in concrete due to neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Irradiation Damages of Structural Materials under Different Irradiation Environments

For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study’s Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.

radiation damage, structural alloys, neutron irrad↗
DOE OSTI2016

Lattice damage and compositional changes in Xe ion irradiated In{sub x}Ga{sub 1-x}N (x = 0.32−1.0) single crystals

Lattice disorder and compositional changes in In{sub x}Ga{sub 1-x}N (x = 0.32, 0.47, 0.7, 0.8, and 1.0) films on GaN/Al{sub 2}O{sub 3} substrates, induced by room-temperature irradiation of 5 MeV Xe ions, have been investigated using both Rutherford backscattering spectrometry under ion-channeling conditions and time-of-flight secondary ion mass spectrometry. The results show that for a fluence of 3 × 10{sup 13 }cm{sup −2}, the relative level of lattice disorder in In{sub x}Ga{sub 1-x}N increases monotonically from 59% to 90% with increasing indium concentration x from 0.32 to 0.7; a further increase in x up to 1.0 leads to little increase in the disorder level. In contrast to Ga-rich In{sub x}Ga{sub 1-x}N (x = 0.32 and 0.47), significant volume swelling of up to ∼25% accompanied with oxidation in In-rich In{sub x}Ga{sub 1-x}N (x = 0.7, 0.8, and 1.0) is observed. In addition, irradiation-induced atomic mixing occurs at the interface of In-rich In{sub x}Ga{sub 1-x}N and GaN. The results from this study indicate an extreme susceptibility of the high In-content In{sub x}Ga{sub 1-x}N to heavy-ion irradiation, and suggest that cautions must be exercised in applying ion-implantation techniques to these materials at room temperature. Further studies of the irradiation behavior at elevated temperatures are warranted.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗