Search NASA⌕ Search

SEARCH · Search NASA

Results for “irradiation effect”

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.

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

At least 19 records
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 OSTI2016

Irradiation effects in tungsten-copper laminate composite

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410-780°C and fast neutron fluences of 0.02-9.0×1025 n/m2, E>0.1 MeV, 0.0039-1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22°C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22°C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile.

Garrison, L. M.↗
DOE OSTI2021

Review of Swift Heavy Ion Irradiation Effects in CeO2

Cerium dioxide (CeO2) exhibits complex behavior when irradiated with swift heavy ions. Modifications to this material originate from the production of atomic-scale defects, which accumulate and induce changes to the microstructure, chemistry, and material properties. As such, characterizing its radiation response requires a wide range of complementary characterization techniques to elucidate the defect formation and stability over multiple length scales, such as X-ray and neutron scattering, optical spectroscopy, and electron microscopy. In this article, recent experimental efforts are reviewed in order to holistically assess the current understanding and knowledge gaps regarding the underlying physical mechanisms that dictate the response of CeO2 and related materials to irradiation with swift heavy ions. The recent application of novel experimental techniques has provided additional insight into the structural and chemical behavior of irradiation-induced defects, from the local, atomic-scale arrangement to the long-range structure. However, future work must carefully account for the influence of experimental conditions, with respect to both sample properties (e.g., grain size and impurity content) and ion-beam parameters (e.g., ion mass and energy), to facilitate a more direct comparison of experimental results.

42 ENGINEERING↗
DOE OSTI2016

Characterization of high energy Xe ion irradiation effects in single crystal molybdenum with depth-resolved synchrotron microbeam diffraction

Microbeam X-ray diffraction experiments were conducted at beam line 34-ID of the Advanced Photon Source (APS) on fission fragment energy Xe heavy ion irradiated single crystal Molybdenum (Mo). Lattice strain measurements were obtained with a depth resolution of 0.7 mu m, which is critical in resolving the peculiar heterogeneity of irradiation damage associated with heavy ion irradiation. Q-space diffraction peak shift measurements were correlated with lattice strain induced by the ion irradiations. Transmission electron microscopy (TEM) characterizations were performed on the as-irradiated materials as well. Nanometer sized Xe bubble microstructures were observed via TEM. Molecular Dynamics (MD) simulations were performed to help interpret the lattice strain measurement results from the experiment. This study showed that the irradiation effects by fission fragment energy Xe ion irradiations can be collaboratively understood with the depth resolved X-ray diffraction and TEM measurements under the assistance of MD simulations. (c) 2015 Elsevier B.V. All rights reserved.

Yun, Di (ORCID:0000000297673214)↗
DOE OSTI2020

Initial development for the introduction of provisions for irradiation effects in Division 5 rules

Heat pipe microreactor concept involves the use of a solid core block to encapsulate the nuclear fuel and the passive extraction of heat from the core block to the heat exchanger block by advanced heat pipes. The core block is a key component of the heat pipe microreactor design. It integrates the functions of reactor vessel, core structural components and fuel cladding of traditional Gen IV advanced reactor designs into one single structure. The effects of irradiation on the structural failure modes are outside the scope of the rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B for metallic Class A components. Methods to account for the irradiation effect were investigated in order to develop design methods that can be applied to the core block design in the heat pipe microreactor. This report presents an initial development of an approach to account for the irradiation effects in the design rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Proton irradiation effects on minority carrier diffusion length and defect introduction in homoepitaxial and heteroepitaxial n-GaN [Proton irradiation effects on minority carrier diffusion length and defect introduction in homoepitaxial n-GaN]

Here, inherent advantages of wide bandgap materials make GaN-based devices attractive for power electronics and applications in radiation environments. Recent advances in the availability of wafer-scale, bulk GaN substrates have enabled the production of high quality, low defect density GaN devices, but fundamental studies of carrier transport and radiation hardness in such devices are lacking. Here, we report measurements of the hole diffusion length in low threading dislocation density (TDD), homoepitaxial n-GaN, and high TDD heteroepitaxial n-GaN Schottky diodes before and after irradiation with 2.5 MeV protons at fluences of 4–6 × 10 13 protons/cm 2 . We also characterize the specimens before and after irradiation using electron beam-induced-current (EBIC) imaging, cathodoluminescence, deep level optical spectroscopy (DLOS), steady-state photocapacitance, and lighted capacitance-voltage (LCV) techniques. We observe a substantial reduction in the hole diffusion length following irradiation (50%–55%) and the introduction of electrically active defects which could be attributed to gallium vacancies and associated complexes (V Ga -related), carbon impurities (C-related), and gallium interstitials (Ga i ). EBIC imaging suggests long-range migration and clustering of radiation-induced point defects over distances of ~500 nm, which suggests mobile Ga i . Following irradiation, DLOS and LCV reveal the introduction of a prominent optical energy level at 1.9 eV below the conduction band edge, consistent with the introduction of Ga i .

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2024

Investigation of ion irradiation effects on mineral analogues of concrete aggregates

Irradiation can cause prominent damage to reactor concrete aggregates leading to amorphization, strength and modulus decrease, radiation induced volume expansion (RIVE) and micro-cracking, which limits their long-term performance. Here, to develop an improved understanding of irradiation effects in concrete, three mineral analogues of concrete aggregates (limestone, marble and quartzite) were irradiated by 5.5 MeV He ions and 13 MeV Ni ions to surface doses of 0.011 displacements per atom (dpa) and 0.23 dpa, respectively, at room temperature. The two different ion species allow irradiation spectrum effects (ionizing and displacive) to be examined. Irradiation induced cracks were observed in He irradiated limestone and marble, and Ni irradiated quartzite. Full amorphization was observed in Ni irradiated quartzite with 14.3 % RIVE, and ∼25 % hardness and modulus decrease, while almost no change was observed in He irradiated quartzite except 4.35 % RIVE, revealing a possible ionization enhanced diffusion effect for high energy light ions. Furthermore, partial amorphization was observed in Ni irradiated marble and limestone matrix with a 12 % hardness decrease in marble while no amorphization was observed for He irradiation with a 20 % hardness increase in limestone matrix. The role of knock-on damage and irradiation spectrum on amorphization, volumetric expansion and mechanical property changes are discussed. Moreover, the onset and critical doses for amorphization and RIVE in quartz are obtained for ion irradiations at room temperature. The dose dependence of RIVE exhibits a delay compared to the amorphization behavior. The superior irradiation resistance of calcite phase compared to quartz phase implies there could be advantages to using calcareous aggregates and lowering the usage of siliceous aggregates for concrete in nuclear power plants for extended operation beyond 60 years. However, other effects such as corrosion, aging and reactions during severe accidents should also be considered, and further investigations are needed.

Amorphous↗
DOE OSTI2016

3D Study of Neutron Irradiation Effects on Tensile Deformation in Fe-based Alloys using High Energy Synchrotron X-rays

Recent development of 3D synchrotron X-ray techniques offers new opportunities for the investigation of nuclear materials. In this study, we use the high energy Xray diffraction microscopy (HEDM), also known as 3DXRD, to investigate the effects of neutron irradiation on microstructural response during tensile deformation in a Fe- 9%Cr alloy. Fe-9Cr has been studied extensively in experiments and modelling as a ferritic model alloy for its response to neutron irradiation. It is found that irradiation-induced hardening and reduction of ductility are significant even at very low doses, and the primary deformation mechanism is channeling. Different from earlier studies, here we focus on grain-level responses to deformation and aim to elucidate the effect of irradiation by employing the far-field HEDM setup (FF-HEDM) to obtain strains, crystallographic orientations, and centers of masses of hundreds of grains in undeformed and deformed states, providing rich 3D information well beyond the TEM and traditional diffraction measurements. It offers a deeper understanding of the structure-property relationship in nuclear materials. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2020

Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility

While additive manufacturing (AM) is an area of increasing interest to the nuclear industry, there are significant research and development needs prior to successful implementation of additively manufactured components into reactors. AM steel behaves differently than traditional wrought material, as has been observed in numerous mechanical testing studies. The nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior (IASCC) of both proton irradiated AM and wrought 316L steel. For IASCC testing, specimens were strained to 4% plastic strain in a simulated boiling water reactor environment (288°C, 0.2 µS/cm). Transmission and scanning electron microscopy, confocal microscopy, and X-ray computed tomography were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This work found that the wrought 316L was more susceptible to IASCC than the AM steel, however, some IASCC was observed in the AM steel when tested with the tensile axis parallel to the build direction. Electron microscopy has shown significantly fewer radiation induced voids formed in the AM steel. While the pre-existing pore network in the AM steel may act as a sink to point defects, this is not expected to be the major factor in the reduced swelling observed in the AM steel. The lower density of radiation induced voids was expected to contribute to the IASCC resistance by decreasing the localization of deformation in the irradiated specimens, however, severity of localized deformation in AM steel, measured in dislocation channel spacing and height, was similar or more severe to that found in wrought. In conclusion, the presence of severe cracking in the wrought specimens did not allow for a direct comparison of localized deformation severities between the AM and wrought steel.

Additive manufacturing↗
DOE OSTI2019

Separate effects irradiation testing of miniature fuel specimens

We report that qualification of new nuclear fuels is necessary for their deployment and requires a thorough understanding of fuel behavior under irradiation. Traditionally, nuclear fuels have been qualified by performing exhaustive integral tests under a limited range of prototypic conditions designed for their specific reactor application. While some integral fuel testing is essential, basic data on behavior and property evolution under irradiation can be obtained from separate effects tests. These irradiations could offer reduced cost, reduced complexity, and in the case of accelerated testing, reduced time to achieve a given burnup. Furthermore, it may be desirable to design test irradiations capable of deconvoluting the myriad effects of burnup, temperature gradients, and other factors inherent to integral irradiation tests. Oak Ridge National Laboratory has developed an experimental capability to perform separate effects irradiation testing of miniature fuel specimens in the High Flux Isotope Reactor (HFIR): the “MiniFuel” irradiation vehicle. The small size (<4 mm 3 ) of the fuel specimens simplifies the design, analysis, and post-irradiation examination. By reducing the fuel mass, the total heat generated inside the experiment vehicle can be dominated by gamma heating in the structural materials instead of fission heating in the fuel. This essentially decouples the fuel temperature from the fission rate, allowing for highly accelerated testing (3X-18X the burnup rate of a typical light water reactor for 235 U enrichments varying from 0.22 wt% to 8 wt%) and an extremely flexible experiment design that can accommodate a wide range of fuel temperatures (~100 °C to >1200 °C), compositions, enrichments, and even geometries without requiring detailed analyses for each fuel variant. In conclusion, this paper summarizes the experiment design concept, evaluates potential applications for specific fuel forms, and briefly describes the first set of experiments on uranium nitride kernels that have been assembled and are currently being irradiated in the HFIR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Neutron Irradiation Effects on Thermal and Mechanical Properties of NITE-SiC

Among the myriad of methods to produce SiC, only a few result in a stoichiometric and crystalline SiC that exhibits stability after radiation damage at temperatures beyond 200 deg. C. High purity SiC processed via chemical vapor deposition (CVD) is the best example of such a material. Another processing method, named nano-infiltration transient eutectic-forming (NITE) SiC, also yields a radiation stable material. It is a fabrication strategy that takes advantage of liquid-phase sintering to yield dense SiC. Specifically, SiC nanopowder feedstock mixed thoroughly with oxide additives are sintered at high-temperature (T<1850 deg. C). This maximum temperature is too low to result in significant creep or grain growth in SiC-based materials, and therefore allows for encapsulation of SiC fibers or SiC coated fuel particles in the NITE SiC matrix. Neutron irradiation effects in NITE-SiC-based materials (composites, monolithic materials, and some with surrogate fuel particles) have been studied to a limited extent. The goal of this study was to examine neutron irradiation effects at two different doses on five different NITE-SiC variants at two different temperatures. Volumetric swelling, degradation of thermal conductivity, and evolution in mechanical properties are reported accordingly.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Comprehensive characterization of the irradiation effects of glassy carbon

Carbon materials have become increasingly diverse, finding applications in high-temperature and high-radiation environments. Glassy carbon, an allotrope known for its exceptional chemical inertness and desirable mechanical properties. However, understanding neutron irradiation effects in glassy carbon has proven challenging, primarily because of its unique nanopore structure. Here, this study presents a highly detailed microstructural characterization investigation of neutron-induced changes in glassy carbon, revealing how changes in nanopore structure and crystallinity impact the irradiation-induced shrinkage. Aberration-corrected scanning transmission electron microscopy (STEM) reveals pore closure that leads to material densification in the irradiated samples. Dimensional analysis combined with comparison to historical data suggest significant length shrinkage to occur. Neutron and in situ electron irradiation experiments suggest that glassy carbon transforms into so-called carbon onions, supporting the concept of shrinkage saturation. Investigating irradiation temperature effects using STEM, electron energy loss spectroscopy, x-ray diffraction, and Raman spectroscopy revealed partial amorphization at 210 °C–230 °C and preserved order in glassy carbon at 860 °C, coinciding with pore closure. Thermal property measurements were also conducted to assess the effects of densification and other changes in the atomic structure of glassy carbon. The results of this study have broad implications in the deployment of glassy carbon to nuclear environments, based around the observed changes in the thermal properties, and demonstrates the operational window for the onset of densification.

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↗