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At least 109 records · Page 6

Self‐Strain Suppression of the Metal‐to‐Insulator Transition in Phase‐Change Oxide Devices

Strongly correlated materials exhibiting phase transitions which can be controlled through external stimuli, such as electric fields, are promising for future computing technologies beyond conventional semiconductor transistors. Devices that take advantage of structural phase transitions have inherent built‐in memory, reminiscent of synapses and neurons, and are thus natural candidates for neuromorphic computing. Of particular interest are phase‐change oxides, which allow for control over the metal‐to‐insulator transition. Here, X‐ray nano‐diffraction structural imaging of micro‐devices fabricated with the archetypal phase‐change material vanadium sesquioxide (V 2 O 3 ) is reported. The devices contain a Ga ion‐irradiated region where the metal‐to‐insulator transition critical temperature is lowered, a useful feature for controlling neuron‐like spiking behavior. Results show that strain, induced by crystal lattice mismatch between the pristine and irradiated material, leads to a suppression of the metal‐to‐insulator‐transition. Suppression occurs within the irradiated region or along its edges, depending on the defect‐distribution and the size of the region. The observed self‐straining effect can extend to other phase‐change oxides and dominate as device dimensions are reduced and become too small to dissipate strain within the irradiated region. The findings are important for phase engineering in phase‐change devices and highlight the necessity to study phase transitions at the nanoscale.

77 NANOSCIENCE AND NANOTECHNOLOGY↗

Characterization of in-situ and ex-situ ion-irradiated additively manufactured 316L and 316H stainless steels

Additively manufactured (AM) 316 stainless steel (SS) differs from its wrought counterpart in its unique dislocation cell structure and the presence of segregation and oxide particles at the cell walls. This work investigated the evolution of the microstructure in laser powder bed fusion (LPBF) 316L and 316H SS under in-situ 1 MeV Kr ion irradiation at 600 °C to 5 dpa, and ex-situ 4 MeV Ni ion irradiation at 300 °C and 600 °C from 0.2 dpa to 10 dpa, with a dose rate for all experiments of 10 -3 dpa/s. The results reveal that the dislocation cell structure results in heterogeneous formation of dislocation loops and voids, particularly at 600 °C, where loops tend to form within the cell interiors while voids form at the cell boundaries. LPBF 316H has a reduced level of swelling compared to LPBF 316L due to prolonged incubation. Energy Dispersive X-ray Spectroscopy (EDS) mapping indicates Ni and Si segregation at void surfaces due to radiation-induced segregation. At 300 °C, where voids are absent, the distribution of dislocation loops and stacking fault tetrahedra appears to be uniform. Dislocation cell structures mostly disappeared by 2 dpa for all conditions in this work. M 23 C 6 carbides were observed in LPBF 316H at 600 °C as early as 0.2 dpa, but not in LPBF 316L. Nanoindentation was performed to obtain the hardness of irradiated materials. In conclusion, this work illustrated the influence of additive manufacturing processes on microstructure evolution under irradiation, revealing the differences as well as the similarities as compared with wrought 316 SS, and the AM-related phenomenon that can potentially occur under neutron irradiation in nuclear reactors.

36 - MATERIALS SCIENCE↗

Final Design of the Material Plasma Exposure eXperiment

The Material Plasma Exposure eXperiment (MPEX) has completed its design phase. MPEX will be a unique facility to investigate plasma material interactions (PMIs) under fusion prototypic divertor conditions in steady state. This includes plasma exposure conditions expected in a fusion reactor divertor. Materials to be investigated will include solids, liquids, and neutron pre-irradiated materials. Electron and ion heating will allow for a large operational domain ranging from erosion-dominated PMI conditions to deposition-dominated PMI conditions. An overview of the final design for all MPEX systems is given. Here, in particular, it is shown how mission-driven project requirements have led to detailed design choices with innovative solutions. Examples are the water-cooled helicon window, the electron cyclotron heating launcher, the target holder and manipulator, and the autonomous decouplers.

Fusion reactor↗

Status memorandum on ORNL support of vendor irradiation capsule design and ASME irradiation code development and new ASTM test standard development activities

This report is in submission of completion of the Level 4 milestone number M4TG-24OR0501111 within the larger Advanced Reactor Technologies Gas Cooled Reactor program at Oak Ridge National Laboratory (ORNL). The focus of this year’s efforts is to continue support of industry needs related to the development of future industry-funded irradiation programs. At ORNL these efforts included the development of a generalized three-dimensional CAD model of an irradiation creep capsules for graphite, cost estimates of the expansion of the Materials Irradiation Facility (MIF), involvement with ASTM D02.F0 “Manufactured Carbon and Graphite Products” and the ASME Boiler and Pressure Vessel Code, and publication of papers supporting these efforts. This report will document the status of these activities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Understanding the limitations and potential of micro tensile testing of tungsten and needs for crystal plasticity modelling

Tungsten is the leading plasma facing material candidate due to its exceptional properties. Understanding the response to neutron irradiation is crucial for the lifetime evaluation of tungsten. The limited space in nuclear reactors and the high levels of radioactivity of the specimens after irradiation are significant barriers to accurately measuring the mechanical properties after neutron irradiation. Testing of micro tensile specimens is one approach to reduce the total amount of irradiated material needed for a set of mechanical testing experiments. However, micro samples are not necessarily measuring the bulk material properties, as size effects produce higher measured mechanical properties than is observed in engineering size counterparts. Determining the minimal sample size for reliable bulk property measurement, combined with modeling efforts, is essential. We conducted room temperature tensile tests on tungsten micro tensile specimens fabricated with focused ion beam, plasma focused ion beam and femto-second laser ablation system to dimension of 2×2×7?µm³, 7×7×18?µm³, and 80×100×233?µm³ (width×thickness×gauge-length), respectively. Elevated tensile testing was performed up to 475°C on 5×5×18?µm³ specimens fabricated with focused ion beam. The smallest specimens exhibited a high degree of ductility and strength, whereas the largest specimens demonstrated behavior akin to bulk tungsten. To investigate the effects introduced by the micro specimen fabrication processes and to obtain the necessary bulk dislocation density for the crystal plasticity model, we employed X-ray diffraction depth profiling. This measurement was performed using different X-ray sources with varying penetration depths. Finally, the potential and limitations of micro mechanical tests for tungsten will be discussed.

36 - MATERIALS SCIENCE↗

A Combinatorial Approach to Testing and Characterization of Irradiated Fuels and Materials

•Combinatorial approaches to testing and characterization of materials is a very powerful tool to help explore some fundamentals of materials science.? •Three topics will be explored to help illustrate this:? •A fitness-for-service assessment of a CANDU component (this is work from my past)? •The use of small-scale mechanical testing of Cr-coated ATF cladding in support of qualification (this is work developed upon joining INL)? •A multi-scale characterization approach of TRISO fuel to investigate fission gas migration and microstructural features (this is work I now get to participate in on joining MFC)? •?

36 - MATERIALS SCIENCE↗

A Workflow to Optimize Fast Neutron Irradiation in A Thermal Neutron Spectrum Test Reactor Leveraging Open-Source Tools

The Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) is one of the key nuclear engineering research and testing facilities within the US Department of Energy (DOE). The ATR is one of few high-power research reactors in the world with different application including accelerated testing of nuclear fuel, materials irradiation in a very high neutron flux environment, and medical radioisotope production [1]. Also, the ATR offers opportunities for testing fast spectrum fission and fusion reactor materials. The key challenges in this area are in further detailing and optimizing a fast spectrum environment within a thermal test reactor. This challenge involves researching, developing, and testing novel concepts for the multiplying of neutron populations into ever higher energy spectra in high flux test reactors like ATR. The main objective of this work is to investigate candidate materials for establishing a fast neutron experiment irradiation in thermal neutron spectrum test reactors which can be accomplished by filtering thermal and epithermal neutrons and boosting fast neutrons at designated irradiation positions. However, adding these filters will render the neutron spectrum and the criticality of the system. The selection of the thickness and material layers should be accomplished by developing an optimization design algorithm that is applicable for ATR to enhance the fast neutron spectrum irradiation utilizing high-fidelity Monte Carlo methods along with advanced machine learning capabilities. This paper presents workflow for design optimization to enhance fast neutron irradiation in the ATR. The workflow leverages open-source tools to develop an algorithm that is viable to ATR and can be leveraged in other reactors. The following sections discuss the development of the experiment design optimization workflow and its application to ATR irradiation positions.

42 - ENGINEERING↗

Photomultiplier window materials under electron irradiation - Fluorescence and phosphorescence

The fluorescence and phosphorescence of photomultiplier window materials under electron irradiation have been investigated using a Sr-90/Y-90 beta emitter as the electron source. Spectral emission curves of UV-grade, optical-grade, and electron-irradiated samples of MgF2 and LiF, and of CaF2, BaF2, sapphire, fused silica, and UV-transmitting glasses were obtained over the 200-650-nm spectral range. Fluorescence yields were determined on these materials utilizing photomultiplier tubes with cesium telluride, bialkali, and trialkali (S-20) photocathodes, respectively. Optical-grade MgF2 and LiF, as well as electron-irradiated UV-grade samples of these two materials, show enhanced fluorescence due to color-center formation and associated emission bands in the blue and red wavelength regions. Large variations in fluorescence intensities were found in UV-grade sapphire samples of different origins, particularly in the red end of the spectrum, presumably due to various amounts of chromium-ion content. Phosphorescence decay with time is best described by a sum of exponential terms, with time constants ranging from a few minutes to several days.

Viehmann, W.↗

Design Report on the MiniFuel Instrumented Test Apparatus for Understanding Radiation Effects

Most nuclear fuels irradiations at Oak Ridge National Laboratory (ORNL) over the past decade have been conducted using MiniFuel—a static capsule design employing subscale fuel specimens to collect separate-effects irradiated fuel performance data. Irradiation conditions for MiniFuel experiments are predicted pre-test using reactor physics, and thermal models are verified post-irradiation via SiC dilatometry and various spectrometry methods. Relevant fuel performance parameters are also observed post-irradiation in a hot cell, thereby providing a single data point for each parameter representing the cumulative effects of the irradiation conditions. Substantially more data can be harvested from a single test and within a shorter duration by instrumenting irradiation vehicles and measuring desired quantities in situ. This report presents the design and analysis of the MiniFuel INstrumented Irradiation Test Apparatus for Understanding Radiation Effects (MINITAURE)—an instrumented test rig based on the separate-effects MiniFuel concept that aims to capture fission gas release (FGR) and thermal conductivity degradation of fuel specimens during irradiation in the High Flux Isotope Reactor (HFIR). MINITAURE will be integrated with the Materials Irradiation Facility (MIF) located in the HFIR building outside the reactor containment. The MIF will act as the instrumentation and control center for the experiment, enabling real-time feedback from in situ sensors and control of irradiation temperatures via a gas delivery system. Two unique capsule designs were developed to capture each phenomenon: the thermal conductivity capsule, which uses a thermopile method to estimate fuel specimen thermal conductivity, and the fission gas release capsule, which will have continuous flowing gas communication to high-purity germanium detectors that are housed in the MIF for monitoring FGR. This report details the reactor physics and heat transfer modeling activities that were used to inform the experiment design and predict capsule performance. It also describes out-of-pile activities conducted to stand up this new capability and verify the measurement techniques. Modeling efforts to date have demonstrated the feasibility of the in situ measurement techniques and supported the development of the MINITAURE assembly configuration. Out-of-pile testing of the thermal conductivity measurement shows promise in capturing relative changes in thermal conductivity. However, significant errors exist in the measured absolute value, posing a need for further refinement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Modified Data Collection And Analysis Codes Of Using Tcm (thermal Conductivity Microscope) To Measure Thermal Conductivity And Diffusivity

The "data collection" basically involves setting up the thermal wave frequency, laser scan distance, and other parameters related to the experimental setup. The modification of this code is minor and the details of this code can be found in the earlier patent ("thermal conductivity microscope"). The "data analysis" instead, replaces the simplified analytical model by a more complete analytical model, and used a "thermoquadruple" method to solve the analytical model. The efficiency is orders of magnitude improved and the accuracy is also better. Meanwhile, the previous model can only handle a two-layer sample structure. The new, complete model can handle materials with multiple layers (any given number), which is necessary to handle post ion irradiated materials.

Hua, Zilong [Idaho National Laboratory (INL), Idah↗

Simulated Galactic Cosmic Ray and Solar Particle Event Radiation Effects on Inflatable Habitat, Composite Habitat, Space Suit and Space Hatch Cover Materials

The effect of particle radiation representative of galactic cosmic ray (GCR) and solar particle event (SPE) radiation on polymeric materials used in deep space mission environments is largely unknown. For NASA, this uncertainty represents an unquantified mission risk. To better quantify this risk, selected polymeric materials used in inflatable habitats, composite habitats, space suits and space hatch covers are irradiated at the Brookhaven National Laboratory, after which changes in relevant end-use properties are measured. The irradiated materials are lightweight candidates having critical functions such as ripstops, permeation barriers, micrometeoroids and orbital debris (MMOD) shield layers and restraint layers.The effect of several types of particle radiation are evaluated. To evaluate GCR effects, high-energy (1 GeV) protons and iron nucleons are used. To evaluate SPE effects, intermediate energy (ca. 20 to 40 MeV) protons are used. In addition, two mission scenarios are evaluated: a Mars mission cycle (space suit materials) and a worst-case 50 year deep space mission cycle (all materials). Lastly, a common polymer (high density polyethylene (HDPE)) is subjected to accelerated aging and radiation exposure to determine if a combined physical aging/radiation effect exists.The level of radiation induced property change observed in this study after exposure of polymers to kGy levels of intermediate and high energy ions is often significant and is quantitatively comparable to the level of change reported for similar materials exposed to MGy levels of low energy protons, electrons and gamma-radiation. This suggests that intermediate and high energy ions encountered in space may cause more damage than low energy forms of radiation due to displacement and linear energy transfer (LET) effects. For these reasons, it is recommended that final approval for mission use be made by the appropriate NASA material and structural review boards to ensure radiation induced changes in candidate materials do not interfere with their engineering function in the intended mission application. Since the materials tested in this report received doses exceeding actual mission doses, the results are considered conservative as a first approximation. However, caution must be exercised when evaluating radiation effects in the absence of other secondary factors know to contribute to degradation. In the addition to radiation type and energy, the results presented in this study show that molecular composition, orientation, stress and physical aging can also contribute to or influence degradation. Other factors such as polymer formulation, secondary radiation effects, stress, exposure to elevated temperature, thermal cycling and exposure to ozone were not investigated here, but must be considered when performing a comprehensive evaluation of a polymeric materials suitability for service in a space radiation environment. Only when the total sum of these factors operating in concert with radiation are accounted for, can accurate assessments of a material's suitability be made.

Jess M. Waller↗

Measurement bias in self-heating x-ray free electron laser experiments from diffraction studies of phase transformation in titanium

X-ray self-heating is a common by-product of X-ray Free Electron Laser (XFEL) techniques that can affect targets, optics, and other irradiated materials. Diagnosis of heating and induced changes in samples may be performed using the x-ray beam itself as a probe. However, the relationship between conditions created by and inferred from x-ray irradiation is unclear and may be highly dependent on the material system under consideration. Here, we report on a simple case study of a titanium foil irradiated, heated, and probed by a MHz XFEL pulse train at 18.1 keV delivered by the European XFEL using measured x-ray diffraction to determine temperature and finite element analysis to interpret the experimental data. We find a complex relationship between apparent temperatures and sample temperature distributions that must be accounted for to adequately interpret the data, including beam averaging effects, multivalued temperatures due to sample phase transitions, and jumps and gaps in the observable temperature near phase transformations. The results have implications for studies employing x-ray probing of systems with large temperature gradients, particularly where these gradients are produced by the beam itself. Finally, this study shows the potential complexity of studying nonlinear sample behavior, such as phase transformations, where biasing effects of temperature gradients can become paramount, precluding clear observation of true transformation conditions.

Crystallography↗

Cosmic-ray exposure history of North Ray and South Ray material

Cosmic-ray exposure ages for two breccias, 68815 and 69935, in the vicinity of South Ray Crater give consistent values of 2.0 plus or minus 0.2 m.y. when measured by the Kr-81-Kr, Na-22-Ne, and particle track methods. These also agree with estimates of exposure ages made by others using microcrater counts on six rocks in the same vicinity, including 69935. It is likely that this represents the age of South Ray Crater. Soils from stations close to South Ray Crater appear to be mature, well-irradiated materials containing little, if any, of a 2 m.y. component. Reviewing various lines of evidence, it is concluded that there are no compelling reasons to believe that so-called South Ray soils contain a large fraction of South Ray ejecta. However, if they do contain such ejecta, this material must have been well irradiated in situ prior to being thrown out in the South Ray event. A chip from a large boulder on the rim of North Ray Crater gives a Kr-81-Kr age of 50.6 plus or minus 3.8 m.y. Similar ages are found by the Na-22-Ne and the particle track methods. The agreement of this result with other data on large North Ray boulders fixes the age of North Ray Crater as 50 m.y.

Behrmann, C.↗

The advanced characterization, post-irradiation examination, and materials informatics for the development of ultra high-burnup annular U-10Zr metallic fuel

U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.

Yao, Tiankai↗

Isotopic Evidence for a Martian Regolith Component in Martian Meteorites

Noble gas measurements in gas-rich impact-melt (GRIM) glasses in EET79001 shergottite showed that their elemental and isotopic composition is similar to that of the Martian atmosphere [1-3]. The GRIM glasses contain large amounts of Martian atmospheric gases. Those measurements further suggested that the Kr isotopic composition of Martian atmosphere is approximately similar to that of solar Kr. The (80)Kr(sub n) - (80)Kr(sub M) mixing ratio in the Martian atmosphere reported here is approximately 3%. These neutron-capture reactions presumably occurred in the glass-precursor regolith materials containing Sm- and Br- bearing mineral phases near the EET79001/ Shergotty sites on Mars. The irradiated materials were mobilized into host rock voids either during shock-melting or possibly by earlier aeolian / fluvial activity.

Rao, M. N.↗

Thermal diffusivity and thermal conductivity of SiC composite tubes: the effects of microstructure and irradiation

Cladding thermal conductivity is an important physical property in assessing the performance of silicon carbide (SiC)-cladded fuels for nuclear reactors. However, there is a significant lack of reliable data, particularly for irradiated materials, because the geometry complicates the measurement. This study investigates the thermal diffusivity of coupons with a curvature, machined from SiC fiber–reinforced SiC matrix composite tubes, with and without neutron irradiation under light water reactor–relevant temperature and dose conditions. The tested materials included full composite and duplex SiC composite tubes. The measurements were conducted using a modern flash diffusivity apparatus. The analyzed area on the specimen during diffusivity testing was reduced for improved measurement accuracy due to sample curvature. Post-irradiation measurements showed that the effects of neutron irradiation on thermal conductivity (e.g. thermal defect resistivity) are different between SiC composite plates versus tubes. The difference was explained by higher matrix density of the tube than the plate. This study provides reliable thermal properties of prototypic SiC composite tubes useful for fuel performance modeling of SiC-based cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of COPVS for High pressure, In-Space, Cryogenic Fuel Storage

Polymeric composite overwrapped pressure vessels (COPVs) provide an attractive material system to support developing commercial launch business and alternate fuel ventures. However to be able to design with these materials, the mechanical behavior of the materials must be understood with regards to processing, performance, damage tolerance, and environment. For the storage of cryogenic propellants, it is important to evaluate the materials performance and impact damage resistance at cryogenic temperatures in order to minimize weight and to ensure safety and reliability. To evaluate the ultimate performance, various polymeric COPV's have been statically burst tested at cryogenic conditions before and after exposure to irradiation. Materials selected for these COPVs were based on the measured mechanical properties of candidate resin systems and fibers that were also tested at cryogenic conditions before and after exposure to irradiation. The correlation of COPV burst pressures with the constituent material properties has proven to be a valuable screening method for selection of suitable candidate materials with resistance to material degradation due to exposure to temperature and radiation.

DeLay, Tom↗

Solute segregation and precipitation across damage rates in dual-ion–irradiated T91 steel

We report that dual-ion irradiations using 5.0 MeV defocused Fe 2+ ions and co-injected energy degraded 2.00 to 2.85 MeV He 2+ ions were conducted on a Fe9CrMo ferritic-martensitic steel T91 to 17 dpa at a damage rate range of 5 × 10 -5 dpa/s to 3 × 10 -3 dpa/s at 445°C, followed by characterization of the microstructure using conventional and scanning transmission electron microscopy. Radiation induced Ni/Si clusters and radiation induced segregation were quantified using energy dispersive X-ray spectroscopy at each condition and were compared with the same material irradiated in the BOR-60 reactor and in the as-received condition. No significant Cr segregation was found at lath boundaries after dual-ion irradiation, while Ni and Si enrichments both decreased with increasing damage rate leading to a sharp decrease in the density of Ni/Si clusters with damage rate. Increased point defect recombination at higher ion damage rates likely reduced the Ni/Si cluster density compared with BOR-60. Although the overall vacancy concentration and diffusion are enhanced by the irradiation damage rate, the lack of time for thermal diffusion and ballistic displacements of solutes are significant limiting factors for Ni/Si cluster formation. This work demonstrates the effect of irradiation damage rate on elemental segregation and clustering when using ion irradiation to simulate reactor irradiation.

36 MATERIALS SCIENCE↗