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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 19 records
All sourcesNASA NTRSDOE OSTI
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 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 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 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 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 OSTI2020

Postirradiation examination from separate effects irradiation testing of uranium nitride kernels and coated particles

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly accumulate burnup during separate effects irradiation testing. In general, the burnup and fuel temperatures measured postirradiation were consistent with the design calculations. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU and was achieved after only 68 effective full-power days of irradiation. The dilatometric evaluation of passive silicon carbide thermometry indicated that the fuel was irradiated at temperatures ranging from 410 to 460 °C. Because the irradiation temperatures and burnup were low, the UN kernels showed minimal fission gas release that was within the range of the expected recoil (athermal) release. While it is possible to measure fuel swelling using x-ray computed tomography, the observed swelling was too small to quantify in this case. Extensive microstructural characterization of the irradiated fuel was performed in this study, and no significant irradiation induced changes were observed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Modeling and design of a separate effects irradiation test targeting fission gas release from Cr-doped UO 2

Fission gas release (FGR) from nuclear fuel during operation can diminish heat transfer properties across the pellet-cladding gap and increase the fuel rod internal pressure, thereby posing a concern to fuel reliability and safety during an accident. Enlarging the fuel grain size, which has been shown to improve fission gas retention, can be achieved by doping the fuel feedstock prior to sintering. In this work, the BISON fuel performance code was used to predict FGR from undoped and chromia-doped UO 2 (referred to as Cr-doped UO 2 ) fuel specimens with different grain sizes and across various temperatures. The BISON models identified the irradiation conditions for which FGR is most significant, and a separate effects irradiation experiment in the High Flux Isotope Reactor (HFIR) was then developed targeting those conditions. Further, the experiment leveraged the MiniFuel irradiation capability at Oak Ridge National Laboratory and consisted of 12 fuel specimens of varying grain size and Cr content. A coupling scheme between BISON FGR results and the ANSYS finite element thermal model used for experiment design was formulated to predict cumulative FGR from each fuel specimen based on expected irradiation temperature histories. The fuel samples were fabricated and characterized as a part of this work, and the fuel compositions modeled in BISON were representative of the specimens used in the experiment. This combined modeling and experimental effort aims to study the effect of fuel grain size and Cr content on FGR and to provide simulated BISON FGR results that can be used for future model validation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2026

Electron beam irradiation effects on bulk metals: a comparative study of polycrystalline versus single-crystalline structures

This study investigates the effects of electron beam (e-beam) irradiation on the mechanical and structural properties of eight bulk metallic samples, comprising both polycrystalline (PC) and single-crystalline (SC) forms of Ni, Cr, V, and Ti. These metals were evaluated as potential candidates for beam exit windows in high-power (MW-class) particle accelerators. The primary objective is to identify metals capable of withstanding the conditions of high-power/MW-class e-beam accelerators and serve effectively as exit windows. Selection criteria were based on each metal’s intrinsic properties, power dissipation capability, and irradiation-induced changes in mechanical behavior, including hardness, elastic modulus, and defect density. Comprehensive characterization was conducted using field-emission scanning electron microscopy (FESEM), atomic force microscopy (AFM), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM–EDS), X-ray diffraction (XRD), high-resolution transmission electron microscopy (HRTEM), selected area electron diffraction (SAED), and nanoindentation, performed both before and after exposure to a ~¿66 kGy dose from a 10 MeV e-beam accelerator. Results revealed that e-beam irradiation induced hardening in PC Ni, whereas PC Ti, commonly used in beam exit windows, exhibited softening. The observed softening in PC Ti is attributed to grain coarsening, elongation, and the formation of twins and twin boundaries, in contrast to the smaller, compressed grains in the pristine (Pr) PC Ti samples, consistent with the Hall–Petch relationship. The stresses due to twinning are small and insignificant in influencing the overall hardening of the PC Ti irradiated sample when compared to the stresses due to the dislocation density. Conversely, SC Ti samples exhibited irradiation-induced hardening. The SC Ti irradiated samples developed additional irradiation-induced modifications in crystallographic texture of (100), (101), (110), (200), (112), (004), and (211) as evidenced from the XRD results, which could probably explain the hardening effect that is caused by irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiation effects on Al 0.3 CoCrFeNi and CoCrMnFeNi high-entropy alloys, and 316H stainless steel at 500 °C

To evaluate the potential of high entropy alloys (HEAs) for nuclear applications, two HEAs, Al 0.3 CoCrFeNi and CoCrFeMnNi, and a conventional Type 316H stainless steel (SS) were irradiated with 1 MeV krypton ions at 500 degrees C up to 1 dpa, and examined in-situ with a transmission electron microscope (TEM). After irradiation, a high density of ordered L1 2 nanoparticles was observed in Al 0.3 CoCrFeNi. In contrast, no phase transformation was observed in CoCrFeMnNi and 316H SS. In the thin foil regions of TEM samples, stacking-fault tetrahedra were observed in the HEAs. In the thick foil regions, interstitial dislocation loops were observed for all alloys, and the areal loop density increased linearly with foil thickness. The Al 0.3 CoCrFeNi had the largest loop size (the lowest density), followed by the CoCrFeMnNi and then the 316H SS. The higher loop nucleation rate in the 316H SS was attributed to carbon content. Finally, the degree of irradiation hardening was slightly lower for the HEAs than for the 316H SS, which is a promising sign for the nuclear application of HEAs at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Multi-energy ion irradiation effects and distinctive features of color center generation in yttria-stabilized zirconia driven by structural discrepancy

Regarding a requirement for inert matrix fuel (IMF) to burn minor actinides and plutonium and high-level waste immobilization, Yttria-stabilized zirconia (ZrO 2 -6.5 wt% Y 2 O 3 ) (YSZ) crystals with cubic phase are selected to comprehensively evaluate the irradiation resistance by irradiation-induced structural discrepancies across multiple energy conditions. Electronic excitation and nuclear collisions reveal distinct color centers, influenced by varying component ratios that reflect their formation mechanisms. Key factors driving ultra-fast structural transitions include collision cascades, pressure waves, and energy dissipation from electronic excitation. In this work, we find that in the electronic energy loss (E ele )-dominant region, internal latent tracks with unique surface nanostructures emerge, distinct from the defect clusters seen in the nuclear energy loss (E nuc )-dominant region, while the threshold for latent track formation and melting are identified. In conclusion, structural discrepancies driven by different energy-loss mechanisms promote the generation of singly ionized and nearest-neighbor doubly ionized oxygen vacancies, which respectively dominate the formation of F⁺ and T-centers, resulting in bandgap narrowing (4.80 eV → 4.73 eV), and furthermore triggering visible light emission (2.17 eV → 2.20 eV) in irradiated YSZ crystals, thus, enabling the design of novel irradiation-tailored material functionalities.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Irradiation Effects on Stability of δ-UZr2 phase in U-50 wt% Zr Alloy

U-50wt%Zr is a candidate metallic nuclear fuel with potential application in light water reactors due to its excellent thermal properties and high radiation tolerance. The Zr-rich UZr fuels possess greater swelling resistance and fission gas release characteristics compared with U-rich UZr fuels. In this current study, the δ-phase U-50wt%Zr is proton irradiated at room temperature to 1 displacement per atom (dpa) to provide insights on phase stability under irradiation conditions. High resolution characterization of Transmission Electron Microscopy (TEM) and Atom Probe Tomography (APT) characterization techniques are used to elucidate microstructural changes due to irradiation. TEM and APT results show highly oriented bcc β-Zr-rich platelet precipitates nucleating adjacent to α-U phases inside the UZr 2 matrix. Formation of this platelet morphology is characteristic of Widmänstatten structure which can be attributed to a variety of factors such as differences in thermal expansion coefficient between the phases, grain size, alloy composition, and cooling rate. The phases present are distinctly different than those observed through in situ annealing, but irradiation accelerates diffusion and phase separation kinetics. These microstructural changes in U-50wt%Zr are different from those achieved by pure thermodynamic or high temperature heavy ion irradiation experiments. Finally, our work, together with previous ones, highlight the necessity to study the U-Zr phase diagram under non-equilibrium thermodynamics conditions to support this material's deployment as a viable nuclear fuel form.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Gamma, electron beam and X-ray irradiation effects on polymers in an advanced bone cement mixer device

The medical device industry has been investigating ways to increase the use of alternatives to cobalt-60 gamma radiation and ethylene-oxide gas sterilization, such as electron beam (E-beam) and X-ray radiation, due to regulatory and market pressures. One impediment to switching to E-beam or X-ray technology for sterilization is the lack of data on the effects of these radiation sources on medical device polymers. To provide such data this work considers irradiation and testing of a common single-use medical bone cement mixing system, the Stryker Advanced Cement Mixer (ACM®), that is composed of seven polymer materials. The ACM® devices considered here were processed to sterilization-relevant doses (15, 25, 50, and 70 kGy) using three radiation technologies: gamma, E-beam, and X-ray. The system and its polymer components were tested for product functionality, as well as mechanical and visual properties to determine how exposure effects may be influenced by radiation technology and dose level. We found that although there were instances of statistically significant differences in effects between the gamma-irradiated products and those irradiated with E-beam and X-ray, those effects were negligible in terms of retained functionality of the product and retained mechanical properties of the polymer components. Overall, results of this study demonstrate that, for of the effects studied, E-beam and X-ray are viable alternatives to cobalt-60 gamma radiation for sterilization of the polymer-based device investigated.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Proton irradiation effects in Molybdenum-Carbide-Graphite composites

The High Luminosity upgrade of the Large Hadron Collider (HL-LHC) has prompted the investigation of novel materials for beam-intercepting devices, and in particular for the collimators responsible for protecting the machine from beam losses. The HL-LHC collimation system will inevitably experience increased levels of radiation damage and undergo changes in their crucial physio-mechanical properties. Graphite-matrix composite materials containing molybdenum carbide particles, along with small amounts of titanium carbide, were developed with the objective of enhanced in-beam performance and tested under proton irradiation. The physical degradation observed in early grades of molybdenum carbide compounds, even after modest proton fluences, has prompted the development of advanced compounds. In this work, we examine the effects of proton irradiation on the microstructural and thermophysical properties of new grades of Molybdenum-carbide-graphite compounds up to fluences of ~2 × 10 20 p/cm 2 . Here, we employ a combination of precision dilatometry and high-energy X-ray diffraction to quantify the dimensional stability and crystallographic phase evolution both pre- and post-irradiation. Our results reveal that these new compounds exhibit superior resilience to radiation damage than their predecessors.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Comparative neutron-irradiation effects on thermal conductivity degradation and dimensional stability of TiC, TiB 2 , and ZrB 2 at 200–1000 °C

Ultra-high-temperature ceramics (UHTCs), including TiC, Ti 11 B₂, and Zr 11 B₂, show great potential for plasma-facing components due to their excellent high-temperature properties prior to irradiation. However, their response to neutron irradiation remains insufficiently understood, limiting robust assessment of their viability for fusion energy applications. Here, this study examines the thermal conductivity, dimensional stability and microstructure of TiC, TiB₂, and ZrB₂ following neutron irradiation at temperatures of 200–1000 °C and fast neutron fluences of 2.0 × 10 25 to 1.1 × 10 26 n/m 2 (E > 0.1 MeV). Lattice swelling measured by synchrotron X-ray diffraction in all three UHTCs was maximized at 200 °C and decreased with increasing irradiation temperature, with no evidence of amorphization observed at 200 °C. Above 600 °C, significant macroscopic volume swelling was observed in irradiated Ti 11 B₂ and Zr 11 B₂, but not in TiC, likely due to cavity formation in the diborides. The post-irradiation thermal conductivity, measured at the irradiation temperature, ranged from 28 to 45 W/m·K, representing a 34–45% reduction relative to the unirradiated material. Notably, neutron-irradiated UHTCs exhibit recoverable thermal conductivity at elevated temperatures, comparable to ferritic–martensitic steels and potentially superior to W when transmutation effects are considered, highlighting promise for shielding or armor plasma-facing components. At 600 °C, both thermal conductivity degradation and lattice swelling saturated at doses exceeding 2–4 dpa.

fusion materials↗
DOE OSTI2021

Irradiation effects in tungsten—From surface effects to bulk mechanical properties

Advanced materials such as tungsten fibre-reinforced composites allow to overcome severe weaknesses of the baseline materials for plasma-facing components — copper and tungsten. The effect of the fusion environment on the mechanical properties of these materials, e.g. the embrittlement by neutron irradiation, plays a key role for the development of future fusion reactors. To simulate this effect, high-energy ions are used as a substitute for the displacement damage by neutrons. We propose the use of very fine tungsten wire as a possibility of studying the influence of irradiation damage on the mechanical properties. This is possible as they allow full-depth irradiation of almost the entire volume despite the limited penetration depth of ions. Geometrical size effects are mitigated due to the nanoscale microstructure of the wire. In addition, similar wire is used in tungsten fibre-reinforced composites. Thus, the investigation of irradiated wire can directly be used for the prediction of the bulk composite properties. For the proof of this concept tungsten wire with a diameter of 16 μm was electrochemically thinned to 5 μm and irradiated with 20.5 MeV W 6+ ions. The mechanical properties were subsequently determined by macroscopic tensile testing. Irradiation to 0.3, 1 and 9 dpa did not lead to a change of the mechanical behaviour. Both strength and ductility, the latter indicated by the reduction of area, were similar to the as-fabricated state.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Accelerated Irradiation Effects Qualification in the AMMT Program

This presentation discusses the ongoing and upcoming efforts to use combined neutron and ion irradiation experiments for materials qualification within the Advanced Materials and Manufacturing Technologies (AMMT) program for the 2024 AMMT program review meeting. The microstructure evolution of additively manufactured 316 L and 316 H stainless steels are discussed, as well as the development of the AMMT challenge problem for this topic.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗