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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Preliminary results addressing material qualification using combined ion irradiation and modeling data

Additively-manufactured (AM) materials have attracted increasing attention in recent years as a new method to make novel and customized components. While AM and conventionally produced materials are compositionally similar, they do possess different microstructures, necessitating assessment of materials produced via AM for their behavior in reactor environments. Some microstructures unique to AM materials, such as compositional micro-inhomogeneity and dislocation cell structures, are of particular importance since they may lead to different radiation performance. The performance of AM materials for advanced nuclear reactor applications is of interest to the Advanced Materials and Manufacturing Technologies (AMMT) program under the Department of Energy Office of Nuclear Energy. The AMMT program aims to demonstrate its new accelerated development and qualification methods via laser powder bed fusion (LPBF) 316 stainless steel (SS). Focusing on material bearing both 316L and 316H specifications, we integrate ion irradiation and modeling. This year, we focus on answering foundational questions related to process variability, alloy chemistry variation, and microchemical segregation. Experimental results provide information and motivate questions to the modeling effort, which aims to develop the ability to model radiation-driven microstructural evolution in additively-manufactured 316 stainless steel under a variety of advanced reactor conditions, including different temperatures, neutron spectra, and fluxes, in a sort of "virtual experiment". We perform in-situ and ex-situ ion irradiations and microstructural characterizations to support the development of AM materials for reactor applications, develop a phase field model of radiation-induced segregation in additively manufactured material with high angle grain boundaries and dislocation cells, and investigate the effect of carbon and chromium content on point defect behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A short review of defect superlattice formation in metals and alloys under irradiation

Irradiation damage drives complex and coupled phenomena in materials at far-from-equilibrium conditions. The self-organization of nanoscale defects in materials under irradiation shows great potential to tailor the physical properties of materials by controlling nanopatterned microstructures. Irradiation-induced gas bubble and void superlattices are two important ordered nanostructures of great scientific interest. Although both types of superlattices have been investigated extensively, a consensus has yet to be reached on their formation mechanisms. In this review article, the current research status of gas bubble and void superlattices in metals and alloys and their characterization, structural stability, and mechanistic modeling are summarized. The fundamental research goals to advance the mechanistic understanding of gas bubble and void superlattices are outlined.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiated Microstructure of Zorita Materials

Reactor baffle plate materials were harvested from the decommissioned Zorita nuclear power reactor after 38 years of service and were characterized with transmission electron microscope. The samples were Type 304 stainless steels and had been irradiated to 0.05, 16.9, ad 48.6 dpa. Faulted dislocation loops, voids, and irradiation-induced precipitates were observed in this material. While dislocation loops were the dominant irradiation defects for all samples, small voids and precipitates were also observed at 16.9 and 48.6 dpa, but not at 0.05 dpa. The size of dislocation loops increased with irradiation dose and saturated below 16.9 dpa. The loop density continued to increase between 16.9 and 48.6 dpa. For voids, both the density and mean size were small, and the volume swelling was negligibly small even at 48.6 dpa. In addition, carbides or G-phase precipitates were detected at 16.9 and 48.6 dpa, and a complex dose dependence was observed. A range of irradiation hardening was estimated at each dose, and the resulting yield strengths were consistent with the previous tensile results measured from the Halden and BOR-60 specimens.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Integrated Approach to Post-Irradiation Examination of Nuclear Materials at Idaho National Laboratory

Idaho national Laboratory (INL) is the U.S. lead national laboratory for the Department of Energy’s Office of Nuclear Energy (DOE-NE), providing much of the nuclear research, development and demonstration capability needed to move nuclear innovation forward to deployment. INL’s Materials and Fuels Complex hosts a unique combination of personnel, facilities and infrastructure and offers the ability to perform post-irradiation examinations (PIE) of nuclear materials spanning multiple length scales. The ability to combine engineering-scale analysis and sub-microscopic characterization provides valuable insights into fuel and structural material behavior and degradation mechanisms. The holistic approach is used to accelerate these materials demonstration and deployment. Selected studies will be presented, highlighting the impact of these techniques on improving fuel reliability, safety, and efficiency, thereby advancing the development of sustainable and advanced nuclear energy technologies.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Effects of He-Ion Radiation on Solid-State Uranyl Nitrate Compounds under Dry and Hydrated Atmospheric Conditions

Radioactive decay of uranium (U) and its related daughter/fission products emit ionizing radiation, including γ (γ) rays and α (α) particles, that result in the formation of radical species and induce chemical reactivity in materials. While radioactivity is inherent to the chemistry of U there are limited studies that detail changes at an atomistic level. Here, this work describes the He-ion radiolysis of four solid-state U(VI) species: [UO 2 (NO 3 ) 2 ]·3(H 2 O) and M[UO 2 (NO 3 ) 3 ] (M = K + , Rb + , Cs + ). These materials were irradiated under different conditions (i.e. closed, open – Ar gas, or open – H 2 O-saturated Ar gas) to further evaluate the impact of water radiolysis on the chemical modification of these materials. Pre- and post-irradiation analyses were conducted using EPR, Raman, and ATR-IR spectroscopy on materials irradiated to 0, 5, 10, and 25 MGy. The results indicated the presence of nitrate radical (NO 3 • ) formation in all solid-state materials with similarities to those observed in γ-radiation studies. Irradiation of [UO 2 (NO 3 ) 2 ]·3(H 2 O) did not show evidence of reactive oxygen species bound to the U(VI) cation under inert conditions; however, surface reactivity was observed for samples irradiated in the H 2 O-saturated environment. Similar chemical changes were observed in the uranyl trinitrato compounds irradiated in the presence of H 2 O vapor and there were observed differences in the reactivity depending on the identity of the alkali cation.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2023

Thermomechanical Properties of Neutron Irradiated Al 3 Hf-Al Thermal Neutron Absorber Materials

A thermal neutron absorber material composed of Al 3 Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al 3 Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4T m . Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al 3 Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al 3 Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al 3 Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of –6.3% for the annealed irradiated 36.5 vol% specimen.

36 MATERIALS SCIENCE↗
DOE OSTI2022

AMTS update for STTR

Advanced nuclear reactor systems currently in development include designs that utilize higher energy neutron spectrums, higher temperatures, and corrosive environments that are quite different from the current fleet of nuclear power plants in use today. As a result, new structural materials must be developed and qualified for use to support these advanced systems. For example, research is currently being conducted on a new class of materials known as multi-principle-element-alloys. These alloys show some promise to be able to meet the structural needs of these challenging reactor systems. However, due to the vast available compositional space, and the limited amount of research conducted thus far, Idaho National Laboratory (INL)sought a way to advanced mechanical testing capability for highly irradiated structural materials to increase both throughput and higher fidelity data analysis through incorporation of digital image correlation. The Automated Mechanical Testing System accomplishes these tasks by automating the testing and incorporating digital image correlation to provide higher fidelity data. INL has now received this system and is currently performing functional testing and training. INL is also performing some modifications to the system to allow for installation into a remote shielded enclosure where testing of irradiated structural materials can be performed. After testing, training, and modifications of the system are complete, INL will install the system into a shielded enclosure and begin using the system to test irradiated structural materials.

36 MATERIALS SCIENCE↗
DOE OSTI2023

AGC-2 Specimen Post Irradiation Data Package Report

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time of flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multiphysics FISPACT-II and TENDL-2019 simulation: neutron-induced damage metrics

Nuclear interactions can be the source of atomic displacement, irradiation-induced defects and transmutation in structural materials. Such quantities are derived from, or can be correlated to, nuclear kinematic simulations of primary atomic energy distributions spectra and the quantification of the numbers of secondary defects produced per primary as a function of the available recoils, residual and emitted, energies. Recoil kinematics of neutral, residual, gas, proton, alpha particle emissions are now more rigorously treated based on recent, complete and enhanced nuclear data parsed in state-of-the-art processing tools. Defect production metrics are the starting point in the complex problem of correlating and simulating the behaviour of materials under irradiation, as experimental information is rare or scattered. Detailed, segregated primary knock-on-atom metrics are now becoming available as the starting point of further simulation processes of isolated and clustered defects in material lattices. This allows more materials, lattices, neutron incident energy ranges, and irradiation conditions to be explored with sufficient data to adequately cover both standard and novel applications and materials: the broader reactor applications landscape. The damage metrics of of materials are systematically explored under typical but different reactor's type environment. The inventory code FISPACT-II combined with the enhanced nuclear data forms of the TENDL-2019 libraries allow one to not only calculate dpa from mostly scattering events but to also properly predict gas production, nuclear heating and transmutation under the same conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Characterization of Intrinsic Fiber Optic Temperature Sensors by In-core and Furnace Testing

Increased research and development in nuclear technology has raised the demand for novel sensors and instrumentation to meet data objectives, and survive in different conditions and environments, beyond conventional light water reactor (LWR) environments. Expediting the deployment of advanced nuclear technologies by developing, demonstrating, and qualifying advanced reactor fuel forms necessitates a deeper understanding of how irradiation affects fuels and materials' performance. To achieve a more comprehensive understanding of fuels and materials performance, researchers require more specialized experiments and measurements. The demand for innovative sensors to support nuclear fuel development arises from the complexity of materials' behavior under irradiation and the challenges of deploying instrumentation in Material Test Reactors (MTRs) for irradiation tests. Additionally, measurements of material properties require integrated measurement systems to characterize thermal properties, mechanical properties, chemistry, and microstructure [1]. Among the potential measurement techniques, optical fiber-based sensors have been identified as potential sensors to measure different physical phenomena such as temperature, strain, pressure, and fluid level. Optical fiber sensors have the capability to provide multi-sensing and multiplexing instrumentation, allowing the measurement of different physical parameters within a single sensor configuration, and transmitting data collected at multiple locations through a single fiber. They offer immunity to electromagnetic interference, electrical passivity, compatibility with various sensing methodologies, and cost-effectiveness. Beyond their widespread use in telecommunications, silica fiber-based instruments are utilized in industrial applications, even at temperatures reaching 300?400°C, such as distributed temperature sensing in oil and gas recovery. The Department of Energy (DOE) is interested in using fiber optics to support fuel cycle development [2]. Fiber optics are an excellent candidate for harsh environment sensing, including sensing at very high temperatures (1900oC for sapphire optical fibers). Distributed fiber optic sensing has been deployed in other harsh environments like coal gasification plants [3]. Distributed strain sensing, which operates similarly to distributed temperature sensing, has been deployed to monitor underground mines and fibers have been imbedded in soil to monitor sink-hole development [4][5]. The application of fiber optic sensors to advanced reactor development is a promising area of research.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Community Data Contribution to M.E.T.A. with ATF-relevant Hydrided Zr cladding (Coated and Uncoated)

Since the aftermath of the Fukushima Daiichi loss-of-coolant accident, accident-tolerant fuel (ATF) claddings have been developed to improve the coping times in such events. However, the mechanical performance of ATF cladding is crucial in ensuring that it does not negatively impact the mechanical integrity during all other stages of the nuclear fuel cycle, and the validity of the existing safe operating margins must be verified. However, due to the cladding’s tube geometry and textured anisotropy, determination of apparent mechanical properties under certain deformation paths is challenging. In the uniaxial hoop direction, for instance, the measured mechanical stresses include frictional forces caused by loading mandrels or varying deformation paths in the sample during traditional ring tensile testing. This experimental difficulty is exacerbated by the specimen size. However, addressing these challenges enables irradiation separate-effects investigations in which the materials can be inserted in reactors like the High Flux Isotope Reactor, and reducing the material consumption of commercially irradiated material allows for further post-irradiation examinations. Despite the advantages of reduced-scale mechanical testing, any drawbacks from new specimen geometries must be evaluated, and uncertainties from specimen preparation, setup, and analysis methodologies must be understood.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Atomic-Scale Characterization of Dilute Dopants in Topological Insulators via STEM–EDS Using Registration and Cell Averaging Techniques

Magnetic dopants in three-dimensional topological insulators (TIs) offer a promising avenue for realizing the quantum anomalous Hall effect (QAHE) without the necessity for an external magnetic field. Understanding the relationship between site occupancy of magnetic dopant elements and their effect on macroscopic property is crucial for controlling the QAHE. By combining atomic-scale energy-dispersive X-ray spectroscopy (EDS) maps obtained by aberration-corrected scanning transmission electron microscopy (AC-STEM) and novel data processing methodologies, including semi-automatic lattice averaging and frame registration, we have determined the substitutional sites of Mn atoms within the 1.2% Mn-doped Sb 2 Te 3 crystal. More importantly, the methodology developed in this study extends beyond Mn-doped Sb 2 Te 3 to other quantum materials, traditional semiconductors, and even electron irradiation sensitive materials.

36 MATERIALS SCIENCE↗
DOE OSTI2023

The microstructure and He + ion irradiation behavior of novel low-activation W-Ta-Cr-V refractory high entropy alloy for nuclear applications

Microstructure and nanohardness of a nearly equimolar W-Ta-Cr-V high entropy alloy (HEA), as well as its irradiation response under He + irradiation, were investigated. The single-phase body-centered cubic nanostructured alloy with a 1 µm thick layer was fabricated on a silicon substrate using a magnetron sputtering method. The HEA film has a complex microstructure consisting of micrometric domains that exhibit internal nanostructure controlled by their crystal orientation. The measured nanohardness of the W-Ta-Cr-V alloy is 13 ± 2 GPa, which significantly exceeds the hardness of nanocrystalline tungsten as a result of the high solid-solution strengthening effect. In order to evaluate the irradiation resistance of the HEA film, the material was irradiated with 200 keV He + ions at room temperature, with two different ion fluences: 1 × 10 16 and 5 × 10 16 ions/cm 2 . Using transmission electron microscopy, a high density of extremely fine He bubbles is observed that were uniformly distributed in the matrix. The increase of He + ion fluence increased the density of bubbles, whereas their size remained at a similar level, which indicates that the damage proceeds by the nucleation of additional He bubbles, not by their growth.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Machine Learning Approaches for Nuclear Material Accounting Data from Irradiation and Reprocessing

We are currently exploring data analysis methods for their ability to strengthen the synthesis and evaluation of information generated within domestic and international safeguard regimes. Safeguard data typically includes rich heterogenous datasets amenable to advanced data analytics methods, such as machine learning. We have converted transactional data containing both numerical and categorical attributes from a domestic nuclear material control and accountability (NMC&A) system into a low-dimensional numerical structure via linear principal component analysis. This data representation allows for global structure discovery via cluster analysis, which can characterize the typical behavior of each of the primary types of transaction events. Furthermore, the structure of the top principal components captures the “typical behavior” of the data and is thus amenable to anomaly detection through statistical hypothesis testing. We explored this capability by generating erroneous permutations of the data and computing the Q-residual quantity increase associated with the information loss when these data are projected into the low-dimensional principal component analysis space representative of typical transactions.Future work will focus on identifying data transformations (e.g., graph networks) that more closely align with the inherent structure of these transactions to explain more salient information and disambiguate the underlying nuclear process—in this case, irradiation and reprocessing—from artifacts of NMC&A system transactional record keeping data.

Drescher, Adam↗
DOE OSTI2021

Vessels Irradiation Experimental Campaign

The purpose of this report is to document the neutron flux field characterization experiments and to provide experiment data to validate radiation transport models. To predict material damage, irradiation experiments expose materials to known radiation fields to correlate the exposure to damage effects. As a part of irradiation experiments, the neutron field in the experiment must be well characterized; this characterization includes effects like room return. Only the neutron component of the 252 Cf sources is considered in theses measurements and calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Testing piezoelectric sensors in a nuclear reactor environment

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE’s FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65x1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 x1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 x1020 nf/cm2 , and Aluminum Nitride is capable of transduction up to at least 8.65x x1020 nf/cm2.

T. Reinhardt, Brian↗
DOE OSTI2020

Method for Evaluating Irradiation Effects on Flow Stress in Fe-9%Cr ODS Using TEM In Situ Cantilevers

Transmission electron microscopic (TEM) in situ mechanical testing has become a widely utilized tool for simultaneously measuring mechanical properties and understanding fundamental deformation mechanisms in irradiated and nuclear materials. Although tensile and compression specimen geometries are amongst the most common, opportunities remain for investigating alternative geometries that could provide unique insights into the plasticity of irradiated materials. This research demonstrates a new TEM in situ cantilever beam configuration. Cantilevers are produced from as-received and proton irradiated (1 dpa, 500°C) Fe-9%Cr oxide dispersion strengthened (ODS) steel. Flow stress is measured using a TEM in situ depth-sensing mechanical testing holder. A 200 MPa increase in flow stress is measured due to irradiation. Size effects arise when the intrinsic (i.e. microstructural) size approaches the extrinsic (i.e. external dimensions) size and can be described using a power law relationship as a function of the material microstructure and cantilever dimensions.

36 MATERIALS SCIENCE↗
DOE OSTI2020

HDG-1 Graphite Preirradiation Data Package Report

This report documents all pre-irradiation examination material-property measurement data for graphite specimens that are going to be used within the first high dose graphite (HDG) -1 irradiation capsule. The two new HDG capsules signify a major change to the AGC Experiment. HDG-1 and HDG-2 will replace the last two Advanced Graphite Creep (AGC) capsules (AGC-5 and AGC-6) which were designed to irradiate graphite at the extreme upper operational temperatures for a very-high-temperature reactor (VHTR) design, 1100°C. These very high temperature AGC-5 and AGC-6 capsules have been repurposed to re-irradiated specimens (from AGC-2, AGC-3, and AGC-4) at the lower temperatures of 600°C and 800°C. HDG-1 will be irradiated at 600°C and HDG-2 will be irradiated at 800°C. By re-irradiating the previous AGC specimens a total maximum neutron dose of around 15 dpa (displacements per atom) can be achieved for all major graphite grades at irradiation temperatures of 600°C and 800°C. Specimens in the HDG-1 capsule are made up of previously irradiated specimens from the AGC-2 capsule and unirradiated specimens prepared for the now discontinued AGC-5 capsule. Utilizing the irradiated specimens, a maximum neutron dose of around 15 dpa is anticipated. These new maximum dose levels will provide irradiated material property data over a total neutron dose range of 1-15 dpa at a temperature of 600°C when combined with the previous AGC-1 and AGC-2 irradiation data. This will provide quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades for use within high temperature reactor designs. Similar to previous AGC test trains, HDG-1 includes the major graphite grades (IG-110, NBG-17, NBG-18, PCEA, and 2114) as well as adding the very fine-grain grade IG-430 which is of interest to the Molten Salt Reactor (MSR) designs. Also new to the HDG-1 capsule are 90 smaller geometry specimens designated as pencil specimens. These specimens take up only one third the space of a standard creep size specimen. This increased number of specimens will enhance property measurement statistics because they will provide 3 times the control specimen data at a position that would otherwise only have a single measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗