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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

An Evaluation and Qualification of U.S.-Based Research Reactors for Irradiation Capabilities Supporting Advanced Nuclear Systems

Irradiation experiments are a prerequisite for evaluating nuclear reactor system designs, analyzing the performance of these systems, and obtaining licenses. Likewise, irradiation facilities are necessary for producing the radioisotopes used in industrial and medical applications. Recent developments in modeling and simulation capabilities and advancements in computational resources have further enabled the design of irradiation experiments for evaluating radiation-induced phenomena and determining nuclear fuel, material, and system design and safety criteria pertaining to both normal and accident scenarios. These computational tools and models require comprehensive experimental datasets acquired under prototypic radiation conditions—for exploring material and system performance under the uniquely harsh environments found in nuclear reactors—to enable verification and validation for qualification and licensing purposes. However, qualification of irradiation experimental facilities, primarily research and test reactors (RTRs), necessitates that their performance be evaluated based on the irradiation environment (e.g. flux, power, testing capabilities) using an appropriate scoring matrix. Although many university campus RTRs are available for research and development (R&D) activities and initiatives, this study focuses on evaluating and qualifying the irradiation facilities (mostly RTRs) within the United States that are suitable for advanced nuclear fuel, material, and system irradiation experiments aimed at establishing operational-performance limits and informing component and fuel designs so as to improve operational efficiencies and mitigate proliferation vulnerabilities, as well as for radioisotope production aimed at multipurpose applications. As a result, the findings of the present study support the acceleration of nuclear fuel and material qualifications, thus hastening new and advanced nuclear energy system demonstrations and radioisotope production efforts by using extended R&D.

irradiation experiment↗
DOE OSTI2024

Phase stability and microstructure of neutron-irradiated substoichiometric yttrium dihydrides

The impact of the neutron-displacement damage on phase stability and microstructure of substoichiometric yttrium dihydrides (YH x , x <2) were investigated to assess their use as solid moderator in high-temperature nuclear reactors. YH x specimens were, thus, subjected to neutron irradiations in the range of 0.1–2 displacements per yttrium atom (dpa-Y) in the temperature range of 536–878°C at the Oak Ridge National Laboratory's (ORNL's) High Flux Isotope Reactor (HFIR). YH x specimens were initially prepared at stoichiometry (H/Y) ratios of 1.69 and 1.83. HFIR-irradiated specimens were characterized by variety of techniques to investigate H retention characteristics including dimensional analysis, optical microscopy, scanning electron microscopy electron back scatter diffraction (EBSD), transmission electron microscopy, thermal desorption spectroscopy (TDS), and high-energy x-ray diffraction (HE-XRD) characterizations. Overall, YH x exhibited notable structural and phase stability under short-term neutron-irradiation, except for the samples with significant silicon carbide (SiC) interaction at high doses and temperatures. Basic dimensional and mass measurements were misleading for accurate assessment of H retention, as confirmed by EBSD phase maps, XRD line profiles, and TDS signals. Thus, it was discussed that a robust H retention metric is needed to assess irradiated hydrides. Further, nanoscale cavities were observed as a result of the neutron irradiation in all samples. Although no clear impact of dose and irradiation temperature was determined, the initial H/Y ratio had an impact on the cavity number density where low H/Y specimens had high-resistance to cavity formation. The Y-vacancy cluster formation at the collision stage of the displacement cascade and their stabilization by H were considered to be the likely underlying mechanisms for the observed cavity microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Irradiation induced void spheroidization, shrinkage and migration in Cu at elevated temperatures: An in situ study

We show understanding the void evolution in irradiation environment is of great interest and significance, as irradiation-induced voids typically lead to pronounced volumetric swelling and degradation of mechanical properties. In situ studies on the irradiation response of nanovoids at elevated temperature remain limited. In this work, we performed systematic in situ 1 MeV Kr ++ irradiations on Cu with nanovoids in a transmission electron microscope up to 350 °C. The in situ studies revealed intriguing void spheroidization, shrinkage and migration. Furthermore, the morphology evolution and migration of nanovoids showed a strong dependence on irradiation temperature and initial void size. Post-irradiation analyses identified defect clusters in the form of stacking fault tetrahedrons, and the remaining large faceted nanovoids. The underlying mechanisms of irradiation-induced void spheroidization and shrinkage were discussed based on phase-field modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effect of heavy ion irradiation dose rate and temperature on α' precipitation in high purity Fe-18%Cr alloy

Cr-rich alpha prime precipitates (CrRP) induce hardening and embrittlement of FeCr alloys, but the kinetics of CrRP formation due to particle irradiation are not well understood. In this study, Fe18wt.%Cr alloy in solid solution state and pre-aged to produce relatively coarse CrRP was irradiated with 8 MeV Fe ions. The irradiation conditions involved two midrange doses of 0.37 and 3.7 displacements per atom (dpa), a wide range of dose rates (10 -5 –10 -3 dpa/s) and temperatures (300–450 °C). The distributions of CrRP after irradiation were studied with atom probe tomography (APT). The critical irradiation conditions to suppress CrRP formation were identified as 300 °C and 10-3 dpa/s; CrRP formation occurred readily at lower dose rates or higher temperatures. From 0.37 to 3.7 dpa, CrRP were observed to slightly grow at 350 °C and strongly coarsen at 450 °C. Specimens with pre-existing CrRP evolved into a similar precipitate distribution as detected after ion irradiation on solution annealed specimens at 300–350 °C to 0.37 dpa, indicating that the precipitate microstructure approaches a quasi-equilibrium for doses <1 dpa. Limited shrinking of pre-existing CrRP was observed after irradiation at 450 °C to 0.37 dpa, indicating a higher recovery rate at this temperature. Additionally, the evolution of CrRP is quantitatively explained by employing corrections to the historic Nelson-Hudson-Mazey precipitate stability model, and a radiation modified precipitation mechanism is proposed to account for the competition between radiation enhanced diffusion and ballistic dissolution which results in the modifications on both size and solute concentration of CrRP.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Irradiation-induced amorphization of UO 2 films prepared by spraying-assisted combustion synthesis

Spraying-assisted combustion synthesis with uranyl nitrate – acetylacetone – 2-methoxyethanol solutions was used to prepare UO 2 films on an aluminum alloy substrate. The tuning of the spraying parameters and annealing temperatures allowed the preparation of UO 2 films with thicknesses varying from 10–300 nm and 5–10 nm UO 2 grain size. High-resolution electron microscopy and X-ray photoelectron spectroscopy showed that increasing the annealing temperature promotes Mg diffusion from the substrate into the films. The incorporation of Mg reduced the overall crystallinity of the films. The irradiation with Ar 2+ ions (1.7 MeV energy and a fluence of 2 × 10 16 ions/cm 2 ) did not degrade the quality of the films. However, the Mg content significantly influenced the irradiation-induced restructuring of the UO 2 films. Irradiated films with low or no Mg content exhibit high crystallinity, and the UO 2 /Al interfacial layer becomes highly porous. Films with higher Mg content are mostly amorphized after irradiation. The origin of irradiation-induced amorphization was related to the formation of Mg y U 1-y O 2±x solid solutions. Chemically complex, pore-free, and amorphous Mg-Al-O film/substrate interfacial layers enable continuous Mg diffusion during irradiation. As a result, the gradual increase in Mg amounts triggers irradiation-induced precipitation of a crystalline MgO-rich phase within the amorphous films.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Lattice response to ambient self-irradiation of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures

Self-irradiation in plutonium is a continuous process that introduces radiation damage ranging from point defects to nanoscale radiation damage as well as decay daughter product impurities that require to be accommodated by the host material. These changes are affecting the structure and properties of the material by lattice swelling, mechanical strengthening, and other alterations. X-ray diffraction and dilatometry studies have shown that the room temperature self-irradiation of δ-phase Pu-Ga alloys (2 and 7 at%Ga) results in swelling of the lattice that saturates after 0.1–0.2 displacement per atom (dpa) of accumulated lattice damage. Here, for a better understanding of these self- irradiation processes, defect accumulation and damage evolution, this time-of-flight neutron diffraction study investigates the impact of ambient self-irradiation effects on the structural evolution of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures (<120 K), associated with a varied history of storage time and thermal treatments. Two key behaviors were observed in the lattice evolution after exposure to cryogenic temperatures with respect to room-temperature accumulated self-irradiation damage. Samples with no self- irradiation damage accumulation under ambient conditions showed swelling of the lattice at the end of the measurement, while samples that accumulated significant self-irradiation damage under ambient conditions showed only a partial recovery of the lattice parameter value upon completion of the measurements (cooling followed by heating back to room temperature). Possible causes of the behavior are proposed, and their associated effects are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Investigation of $\gamma$-irradiated polyvinylidene fluoride and its acute toxicity

Polyvinylidene fluoride (PVDF) is an industrial polymer with many applications. The EPR spectrum of polyvinylidene fluoride (PVDF) irradiated by γ-rays at low temperature (-196°C) provides evidence for the presence of radicals formed by loss of a fluorine atom from PVDF, and polyenyl radicals. Polyenyl radicals dominate in the polymer irradiated at room temperature. Peroxide radicals formed as a result of oxidation of primary free radicals by atmospheric oxygen are observed in the EPR spectrum of a polymer irradiated in air, or when air is introduced to a polymer irradiated in a vacuum. Carbonization and oxidation of γ-irradiated PVDF were observed in the XPS spectrum. A model for polymer degradation based on correlated G3(MP2) molecular orbital theory electronic structure calculations is described in this report. The presence of HF in the products and a low yield of products with the composition CxFyHz indicate the absence of noticeable thermal destruction of the main polymer chain upon heating to 220°C of both the initial and γ-irradiated PVDF. Although any HF released during the PVDF dehydrofluorination would be toxic, a study of the acute toxicity of γ-irradiated PVDF showed that the polymer does not exhibit toxicity after a single intragastric route of administration to mice.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt.% zirconium (U-10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were determined by image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (~ 8 x 10 5 µm 3 ) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U-10Zr fuel region, previously irradiated in the Fast Flux Test Facility, and historic U-10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II) was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

In-situ ion irradiation study of alloy 709 stainless steels with different processing histories

Alloy 709 stainless steel (A709 SS) has gained an increased interest in nuclear applications due to its advanced properties at high temperatures over the widely used Type 316 SS. However, how this alloy behaves under irradiation with concurrent environmental factors, such as mechanical deformation and/or thermal annealing, has not been studied. In this work, we used in-situ ion irradiation inside a transmission electron microscope (TEM) to investigate the effects of irradiation on the microstructures of A709 SS samples with different prior processing histories, including as-received, annealed, and creep-deformed. It was found that, compared to room temperature irradiation, 600 degrees C irradiation led to larger dislocation loop sizes and lower loop densities. Furthermore, with 600 degrees C irradiation, the creep-deformed sample, which had the highest precipitate density and the highest dislocation density, had the smallest loop size and the lowest loop density compared to others. This observation has been rationalized with quantitative evaluation of the sink strength factors in each sample. This study shows that in-service deformation can significantly affect the irradiation performance in structural materials at reactor operating conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2023

High temperature annealing of irradiated nuclear grade graphite

Previous work has shown that the material properties of nuclear grade graphite are substantially affected by the atomic and microstructural changes that occur during neutron-irradiation. The parameters that play a role in these changes are irradiation dose, temperature, graphite composition and the initial microstructure of the graphite. Understanding the details of how these material changes occur and exactly what the changes consists of will enable the prediction of the material property changes as a function of these variables. Other work has shown these atomic level and microstructural changes can be healed or annealed out by raising the irradiated graphite above its irradiation temperature. Here in this work, experiments were carried out to investigate how the properties of irradiated graphite recover when heated. By showing property recovery as a function of annealing temperature or energy, insight is provided into the type of damage that occurred during neutron-irradiation. The data presented here shows recovery of thermal diffusivity, coefficient of thermal expansion, Young's modulus and electrical resistivity between annealing temperatures of 500 °C and 2380 °C. Graphite grades NBG-18, IG-110 and PCEA are considered that were both stressed and unstressed during irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A spatially-resolved model of neutron-irradiated tungsten coupling stochastic cluster dynamics and finite deformation plasticity

Structural materials used in nuclear reactors face severe degradation in mechanical properties, such as hardening and embrittlement. At the microscopic scale, this occurs due to creation and accumulation of irradiation-induced defects and their interaction with system dislocations. Although techniques exist which can model evolution of irradiation defects, for instance kinetic transport theory-based models, their interaction with mechanical deformation of the bulk material has not been investigated extensively. In this work, we demonstrate a novel spatially-resolved multiscale coupling between microscopic irradiation defect evolution, modeled using Stochastic Cluster Dynamics (SCD) and macroscopic mechanical deformation modeled using a finite-deformation plasticity model. SCD is used to determine the statistically averaged defect cluster spacing, dependent on operating conditions such as irradiation dose and temperature. This acts as an initial condition that governs the critical resolved shear stress of dislocation glide in the macroscopic plasticity model. This framework is used to predict mechanical behavior in post-mortem test of irradiated Tungsten samples, which has found its importance as structural material used in nuclear reactors. The results obtained using the coupled approach are in good agreement with experimental data of uniaxial tension tests. The model is able to capture the effect of temperature and irradiation dose on the material hardening. Two methods are proposed to estimate hardness – using Tabor's Law relating uniaxial yield stress to hardness and from flat-punch simulations. The results are in reasonable agreement with hardness data from micro-indentation experiments of irradiated Tungsten samples. Finally, the model is also able to reveal microstructural details such as spatial variation in defect density and local stress.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

A mechanism-based quantitative multi-scale framework for investigating irradiation hardening of tungsten at low temperature

We present the development and validation of a mechanism-based multi-scale modeling framework to quantitatively link the irradiation defect evolution kinetics at the microscopic scale to the macroscopic yield strength and flow stress evolution of tungsten irradiated at low temperature (0.08 < T/T m < 0.18, where T m is the melting point and is equal to 3673 K). The mechanism-based strength model, proposed as the superposition of thermal softening and modified dispersed barrier hardening, is developed to understand the underlying strengthening mechanisms. The thermal softening exponent as defined in the Johnson-Cook model is obtained by fitting unirradiated yield strength at different temperatures. A set of irradiation-induced defect kinetics equations with the thermodynamics parameters derived from the atomic calculations reported in the literature is used to determine the densities and sizes of defect clusters at the meso-scale, and the predicted irradiation damage characteristics are in reasonable quantitative agreement with experimental data from literature. The effects of irradiation condition (temperature and irradiation dose) and test temperature on yield strength are quantitatively predicted and compared with experimental measurements from the literature. Here, the predicted irradiation hardening and strain hardening are compared with experimental data as model validation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Experiment design for the neutron irradiation of $\mathrm{PM-HIP}$ alloys for nuclear reactors

Here, this article describes the design of an Advanced Test Reactor (ATR) drop-in neutron irradiation experiment aiming to directly compare the performance of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP) against conventional casting or forging. There is growing interest in PM-HIP alloys for nuclear applications because of their microstructural uniformity, superior mechanical properties, and reduced dependence on welding and machining, compared to cast/forged alloys. Nuclear code-qualification of PM-HIP alloys requires neutron irradiation testing to demonstrate performance under relevant conditions. In this experiment, six nuclear structural alloys were irradiated: Ni-based alloys 625 and 690, Grade 91 ferritic steel, SA508 pressure vessel steel, and 304L and 316L austenitic stainless steels. The experiment is assembled into seven capsules in four test trains and irradiated in three ATR inboard A positions. Both the PM-HIP and cast/forged versions of each alloy were irradiated under nearly identical conditions for comparative purposes, to target doses of 1 ± 0.2 and 3 ± 0.2 dpa at temperatures of 300 ± 50 °C and 400 ± 50 °C. A thorough description of the experiment design and thermal, structural and neutronic analyses performed to ensure the targeted irradiation conditions are met is provided. Specimens were configured as small disks, compact tension specimens and tensile bars to facilitate post-irradiation examination (PIE) that will include mechanical testing, microstructure characterization, and fracture toughness testing. Given the considerations for ASTM standardized mechanical testing, comparative fluence and temperature across specimen pairs, and comprehensive PIE planning herein, this work serves as a template for future nuclear materials qualification experiment designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Evaluation of clear-sky and satellite-derived irradiance data for determining the degradation of photovoltaic system performance

Knowing the degradation in performance of a photovoltaic (PV) system over time is important for estimating the lifetime energy produced and the financial return. A key parameter for normalizing performance and determining degradation is the plane-of-array (POA) irradiance. Because accurate long-term POA measurements are not always readily available, three methods of providing irradiance data for determining the degradation rate of PV systems were evaluated—a method using irradiance data modeled with the Ineichen clear-sky model and monthly Linke turbidity coefficients, a method using the supplemental clear-sky irradiance data from the National Solar Radiation Data Base (NSRDB), and a method using the NSRDB solar irradiance data for both cloudy and clear-sky conditions (all-sky). The irradiance data from the three methods were evaluated using measured irradiance data from 1998 through 2018 for the seven-station SURFRAD network and for 3-, 5-, and 10-year periods that might be used for evaluating PV system performance. Only the two clear-sky methods for the 10-year periods had less uncertainty with respect to determining PV system degradation than the expected median degradation rate for PV systems of -0.5%/year to -0.6%/year. Shorter periods and the all-sky method had larger uncertainties, making their use questionable for determining the degradation rates of PV systems.

14 SOLAR ENERGY↗
DOE OSTI2023

Interinstitutional Study of the New EURO-GANEX Process Resistance by Gamma Irradiation Test Loops

As part of the homogeneous actinides recycling strategy, the EURO-GANEX process is one of the most promising options to achieve the goal of minor actinides recovery. However, EURO-GANEX also has various drawbacks that are being addressed. Improvements made to EURO-GANEX system have resulted in the emergence of the so-called New EURO-GANEX process, where the composition of the solvent has been modified by replacing TODGA and DMDOHEMA with cis-mTDDGA in the organic phase and SO3-Ph-BTP with PyTri-Diol in the aqueous phase in order to resolve important issues. The objective of this work is twofold: evaluate the gamma radiolysis resistance of the new EURO-GANEX process by dynamic irradiation conditions simulating the three main steps of the process and validate the design of CIEMAT Náyade, CEA Marcel and INL irradiation loop devices since each of them mimics different aspects of the real process. Náyade and the INL loops could irradiate together the organic and aqueous phases, whereas in the CEA loop, the irradiated solvent is recycled continuously inside a platform with several stages of mixer-settlers containing aqueous flows simulating the three main steps of the process. The extraction performances and changes in the composition of the solvent have been analysed during the irradiation experiment by different techniques: gamma spectrometry and ICP-MS/OES for cations or tracers extraction, and HPLC-MS to quantify the degradation and evaluate the degradation compounds. Additionally, in spite of some differences between the three-irradiation facilities, this inter-institutional study shows that they are three comparative tools, which provide results on the stability towards radiolysis of a liquid-liquid extraction system. Favourable extraction results for the different steps are obtained according to the static irradiation studies found in literature. However, the degradation of cis-mTDDGA is appreciable leading to degradation compounds, some of which form precipitates and produce important changes in viscosity, important aspects that must be addressed prior to the successful industrial application of the new EURO-GANEX process.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Effects of temperature and dose rate on ion-irradiated γ-LiAlO 2 pellets

Defect accumulation and microstructural evolution during ion irradiation at elevated temperatures are governed by competing processes of defect production, driven by the dose rate, and defect recovery, controlled by diffusion, interaction, and annihilation. Here, this study investigates the effects of irradiation temperature and the dose rate on microstructural evolution, deuterium retention, and lithium volatilization in γ-LiAlO 2 pellets subjected to sequential He + and D + ion irradiation. Experiments were performed to a total fluence of 3 × 10 17 (He + + D + )/cm 2 at 623, 673, 723, and 773 K with an average He + dose rate of 7.7 × 10 −4 dpa/s, and to 2 × 10 17 (He + + D + )/cm 2 at 773 K with dose rates of 6.8 × 10 −5 , 2.9 × 10 −4 , and 7.3 × 10 −4 dpa/s. At 623 K, the microstructure was dominated by cavities and fractures with no observable precipitate formation, while small precipitates emerged at 673 K. Increasing the irradiation temperature to 723–773 K promoted the formation of larger, faceted LiAl 5 O 8 precipitates, and surface amorphization, accompanied by pronounced lithium depletion and H–D isotopic exchange. At 773 K, medium and high dose rates produced an amorphized surface layer over a crystalline subsurface containing LiAl 5 O 8 precipitates and blisters at the crystalline–amorphous interface, whereas low-dose-rate irradiation preserved surface crystallinity with cavities distributed in the matrix, around precipitates, and along grain boundaries. Precipitate morphology was anisotropic with limited size dependence on the dose rate. These results elucidate the coupled effects of temperature and the dose rate and demonstrate that sequential He + and D 2 + irradiation at 773 K reproduces key microstructural features and H isotope behavior observed in neutron-irradiated γ-LiAlO 2 at 573 K.

dose rate effects↗
DOE OSTI2024

Nanoscale Consequences of Irradiation Investigated by RAD-AFM (Final Report)

To assess the feasibility of measuring material property changes induced by ion irradiation one pristine and three legacy HT-9 samples were sectioned and polished. The polish protocol was first optimized using a pristine unirradiated HT-9 sample before polishing the ion irradiated samples as the spot size was small and repeated polishing could remove the ion irradiated region entirely. AFM tapping mode was used to assess the surface quality after each polish. AFM based hardness mapping was optimized for new diamond tips specifically for use with the HT-9 samples to compensate for differences in grain size, precipitates, and heterogeneity of hardness relative to past samples. In addition, a hardness calibration protocol was developed using NIST copper, nickel, and steel reference materials. Co-located SEM-EDS and AFM analysis was used to verify the ion irradiated edge using the silver/tungsten coating as a fiducial mark. Once benchmarked, the hardness across the ion irradiated cross section, a comparison of bulk hardness according to neutron irradiation, and the hardness of the ion irradiated surface edge was measured.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A Novel Microshear Geometry for Exploring the Influence of Void Swelling on the Mechanical Properties Induced by MeV Heavy Ion Irradiation

Small disks are often the specimen of choice for exposure in nuclear reactor environments, and this geometry invariably limits the types of mechanical testing that can be performed on the specimen. Recently, shear punch testing has been utilized to evaluate changes arising from neutron irradiation in test reactor environments on these small disk specimens. As part of a broader effort to link accelerated testing using ion irradiation and conventional neutron irradiation techniques, a novel microshear specimen geometry was developed for use with heavy-ion irradiated specimens. The technique was demonstrated in pure Cu irradiated to 11 and 110 peak dpa with 10 MeV Cu ions. At 11 peak dpa, the Cu specimen had a high density of small voids in the irradiated region, while at 110 peak dpa, larger voids with an average void swelling of ~20% were observed. Micropillar and microshear specimens both exhibited hardening at 11 dpa, followed by softening at 110 dpa. The close alignment of the new microshear technique and more conventional micropillar testing, and the fact that both follow intuition, is a good first step towards applying microshear testing to a wider range of irradiated materials.

36 MATERIALS SCIENCE↗