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Chancey, Matthew Ryan

Publications and source records attributed to Chancey, Matthew Ryan.

16 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Study of Helium Irradiation Effect on Al6061 Alloy Fabricated by Additive Friction Stir Deposition

Additive friction stir deposition (AFS-D) is considered a productive method of additive manufacturing (AM) due to its ability to produce dense mechanical parts at a faster deposition rate compared to other AM methods. Al6061 alloy finds extensive application in aerospace and nuclear engineering; nevertheless, exposure to radiation or high-energy particles over time tends to deteriorate their mechanical performance. However, the effect of radiation on the components manufactured using the AFS-D method is still unexamined. In this work, samples from the as-fabricated Al6061 alloy, by AFS-D, and the Al6061 feedstock rod were irradiated with He+ ions to 10 dpa at ambient temperature. The microstructural and mechanical changes induced by irradiation of He+ were examined using a scanning electron microscope (SEM), energy-dispersive X-ray spectroscopy (EDS), transmission electron microscopy (TEM), and nanoindentation. This study demonstrates that, at 10 dpa of irradiation damage, the feedstock Al6061 produced a bigger size of He bubbles than the AFS-D Al6061. Nanoindentation analysis revealed that both the feedstock Al6061 and AFS-D Al6061 samples have experienced radiation-induced hardening. These studies provide a valuable understanding of the microstructural and mechanical performance of AFS-D materials in radiation environments, offering essential data for the selection of materials and processing methods for potential application in aerospace and nuclear engineering.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Proton Quenching in Rare-Earth Inorganic Scintillators: GAGG:Ce and YSO:Ce

Scintillator detectors are an integral component of radiation detection systems for a variety of applications such as medical imaging, accelerator diagnostics, and space science. Typically, a scintillator detector’s response is characterized using gamma sources to understand the detection response to different types of radiation, including charged particle detection. However, there exists a nonlinearity of the amount of light produced from an incident gamma ray of specific energy and the light produced from an incident charged particle of the same energy. This important effect, known as quenching, must be accounted for to interpret energies from charged particles incident on detectors. In this article, we present results of quenching parameterization for two types of cerium-doped inorganic scintillators, Y2SiO5:Ce (YSO:Ce) and Gd3Al2Ga3O12:Ce (GAGG:Ce). We measured the light output from incident proton energies from 1 to 25 MeV using a 3-MV tandem accelerator and two reactions: Au(p,p)Au and 3He(d,p)⁴He. Using gamma-ray sources to calibrate the detectors, we compared the measured electron-equivalent energy versus the incident energy expected. Using an adaptation of the Birks semi-empirical formula, we extracted the Birks parameter (kB) to understand quenching. For one of the GAGG:Ce samples, the kB parameter of 0.0072 [g cm-2 MeV-1] is comparable to a similar study where the value of kB was 0.0065 [g cm-2 MeV-1]. For YSO:Ce, no other kB values were found in the literature. Three different types of GAGG:Ce were used to collect measurements of kB as a function of dopant concentration.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Transition in helium bubble strengthening of copper from quasi-static to dynamic deformation

Damage from low-temperature irradiation and the subsequent degradation of materials performance pose significant challenges for the storage of radioactive materials and for peripheral components in some nuclear reactor designs. Fully understanding the mechanical behavior of such materials requires test data for strain rates in both the quasi-static (< 10/s) and dynamic (>> 10/s) regimes. While dynamic testing has generally been avoided in the past for neutron irradiated (contamination concerns) and ion irradiated (insufficient volume) materials, surface-sensitive Richtmyer-Meshkov instability (RMI) tests were used in the present work to overcome these limitations. Here, nanopillar compression, nanoindentation, and RMI testing data from a helium implanted surface layer (~10 µm thick) were compiled to explore the effects of helium bubbles on the materials strength of high-purity copper at strain rates of 0.001/s – 10 8 /s. While nano-mechanical testing revealed increases in yield strength and hardness with increasing helium dose from 1000 to 4000 appm He, RMI indicated no significant changes in strength as compared to unimplanted copper. Here, this discrepancy in behavior was rationalized through a combination of recent literature and follow-on molecular dynamics (MD) simulations, leading to the conclusion that the nanoscale helium bubbles acting as dispersed barriers to dislocation motion at quasi-static strain rates collapse under shock loading and cease to be effective barriers at high strain rates.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Micromechanical properties of spherical and facetted He bubble loaded copper

Exploring new irradiation resistant materials requires understanding their mechanical responses to irradiation. Resistance to helium bubble formation and understanding bubble effects on the mechanical response of candidate materials are crucial factors to qualify materials as irradiation resistant. Here, in this paper, we explore the effect of spherical and facetted helium bubbles on the mechanical response of copper via in-situ micromechanical tensile testing at room temperature. Bubble formation and shape effects on strength and ductility, and their behavior on grain boundaries are discussed and compared to literature. Loading Cu with helium bubbles is shown here to increase strength but decrease ductility.

42 ENGINEERING↗
DOE OSTI2022

Interface effect of Fe and Fe 2 O 3 on the distributions of ion induced defects

The stability of structural materials in extreme nuclear reactor environments—with high temperature, high radiation, and corrosive media—directly affects the lifespan of the reactor. In such extreme environments, an oxide layer on the metal surface acts as a passive layer protecting the metal underneath from corrosion. To predict the irradiation effect on the metal layer in these metal/oxide bilayers, nondestructive depth-resolved positron annihilation lifetime spectroscopy (PALS) and complementary transmission electron microscopy (TEM) were used to investigate small-scale defects created by ion irradiation in an epitaxially grown (100) Fe film capped with a 50 nm Fe 2 O 3 oxide layer. In this study, the evolution of induced vacancies was monitored, from individual vacancy formation at low doses—10 –5 dpa—to larger vacancy cluster formation at increasing doses, showing the sensitivity of positron annihilation spectroscopy technique. Furthermore, PALS measurements reveal how the presence of a metal–oxide interface modifies the distribution of point defects induced by irradiation. TEM measurements show that irradiation induced dislocations at the interface is the mechanism behind the redistribution of point defects causing their accumulation close to the interface. Finally, this work demonstrates that the passive oxide layers formed during corrosion impact the distribution and accumulation of radiation induced defects in the metal underneath and emphasizes that the synergistic impact of radiation and corrosion will differ from their individual impacts.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effects of helium cavity size and morphology on the strength of pure titanium

In this study, high-purity α-titanium was implanted with helium to observe cavity morphology – size, number density, and form (i.e., bubbles vs. voids) – effects on materials strength. Increasing implantation temperatures lead to an Arrhenius-type (i.e., exponential) increase in cavity size, a transition from spherical bubbles to faceted voids, and a marked increase in strength (~20–50%). Implanted samples deformed uniformly in contrast to nano-compression of unimplanted or alloyed titanium, but pillars with the largest faceted voids developed a type of bulging localized deformation. Affected voids were repeatedly sheared, becoming rough and shrunken. Dislocation-cavity interactions and the role of helium content within cavities are considered.

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 OSTI2021

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe 2+ up to 600 peak displacement-per-atom (dpa) at 450 °C. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. Furthermore, the comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Comparison of void swelling of ferritic-martensitic and ferritic HT9 alloys after high-dose self-ion irradiation

In this work, ferritic-martensitic (F/M) HT9 along with fully ferritic HT9 specimens were irradiated using self-ions to 600 peak displacement-per-atom (dpa) at 450 °C. To investigate and minimize the carbon contamination effect on void swelling, two identical F/M HT9 specimens were prepared: one without a coating and the other one with a tungsten and silver coating on the surface prior to irradiation. The specimen without a coating showed carbide formation in the matrix, resulting from carbon contamination during irradiation, with no voids observed. The coated sample did not form matrix carbides, and voids were observed within the irradiated region. The same coating was applied to the ferritic HT9, and void swelling results of both the F/M HT9 and the ferritic HT9 alloys were compared after irradiation. The F/M HT9 exhibited a maximum of 2.6% swelling, while the ferritic HT9 showed a maximum of 0.8% swelling at a 400–500 nm depth, equivalent to 334 local dpa. The results were correlated with the instability of carbides in the ferritic HT9 under irradiation. Carbides in the F/M HT9 and the ferritic HT9 are both M23C6, but the matrix carbides in the ferritic HT9 were less stable under the irradiation, compared to the ones on the grain boundaries in the F/M HT9. The study shows the impact of chemistry on the void swelling which overpowers the microstructural factors such as grain size and phase.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Influence of solidification structures on radiation-induced swelling in an additively-manufactured austenitic stainless steel

Metal additive manufacturing offers potential advantages for producing structural materials, such as austenitic stainless steels, in nuclear power systems. However, the microstructure developed during metal additive processing is notably different from the one developed in conventional processing, and the influence of the microstructural differences on performance in radiation environments has not been fully quantified. Using heavy ion irradiation and transmission electron microscopy, the radiation-induced swelling response of a laser powder-bed fusion-manufactured austenitic stainless steel was investigated at high doses. The influence of solidification-induced dislocation and precipitate structures was studied by comparing the radiation-induced swelling response of a 316L stainless steel in three microstructural states: as-fabricated, solution annealed, and fully recrystallized. Void swelling was approximately twice as pronounced in the as-fabricated state compared to post-processed states. In the framework of the rate theory for radiation effects, the higher swelling in the as-fabricated state can be explained by the strong sink bias for interstitial point defects exerted by the intermediate density of pre-existing dislocations. Void swelling was inhibited in the vicinity of pre-existing precipitates, but the density of precipitates in the as-fabricated material was not enough to compensate for the increase in swelling caused by dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Nanomechanical properties of pristine and heavy ion irradiated nanocrystalline tungsten

Nanocrystalline materials are postulated as strong candidates for nuclear materials applications due to their high irradiation tolerance manifested by their high grain boundary density. Their mechanical properties before and after irradiation, however, are a concern. Nanoindentation, micropillar compression and cantilever bending tests were performed on nanocrystalline tungsten films (grain size less than 35 nm) prior to and post irradiation at high temperature (1050 K) with heavy ions. Their performance under different stress states corresponding to the different tests was evaluated. Characterizing un-irradiated and irradiated nanocrystalline materials was shown to be best performed using a combination of different micromechanical tests.

36 MATERIALS SCIENCE↗
DOE OSTI2018

High Dose Ion Irradiation Testing on Improved HT-9 Alloys

HT9 alloys having various nitrogen contents were deformed up to 50% engineering strain by plane strain compression. Dislocation density and characteristics were determined after neutron diffraction using line profile analysis. These samples having low and high nitrogen contents were irradiated at 300 °C up to 23 peak dpa. In this study, effect of plane strain deformation on the radiation response of low and high N HT9 alloys has been investigated using both transmission electron microscopy and nanoindentation. It has been found that compression results in a considerable increase in both dislocation density and edge type dislocations. Both TEM and nano-indentation studies indicate that plane strain deformation results in improved radiation resistance in high N HT9 alloys. It has been also found that further deformation beyond 10% compression does not improve the radiation resistance further. On the other hand, deformation does not have a clear effect on the radiation resistance in the case of low N HT9 alloys. Together with that, ion irradiations at 300 °C results in ~2 times higher hardening in low N alloys compared to high N alloys.

36 MATERIALS SCIENCE↗