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At least 1,783 records · Page 99
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Microstructure-Level Modeling of Stage 3 Fission Gas Release in UO 2 Fuel

A new phase-field model of fission gas bubble evolution was developed and applied to gain an improved understanding of the microstructure-level processes leading to fission gas release from nuclear fuel and to inform improvements to the fission gas release model used in BISON. The phase-field model simulates the evolution of multiple fuel grains and intergranular bubbles simultaneously, tracks the local concentration of vacancies and gas atoms, and includes the effect of hydrostatic gas pressure on the surrounding fuel matrix. This represents the most realistic microstructure-level model of fission gas bubbles developed to date. The model was used to simulate the growth of grain face and grain edge bubbles in a hexagonally periodic 3D grain structure, which contains multiple grain boundaries and triple junctions. The coverage of grain boundaries and triple junctions by fission gas bubbles was calculated as a function of time. Preliminary results suggest that grain edge bubbles do not release a significant amount of gas prior to gas release by grain face bubbles, which supports the assumptions of the fission gas release model currently used in BISON. However, the simulations should be improved by including a physically realistic number of grain edge bubbles in the initial conditions before definitive conclusions can be drawn.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Complete Status Report Documenting the Processing, Microstructure and High-Temperature Mechanical Properties of the Nanostructured Ferritic Alloy OFRAC

This report provides a summary of the new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), that was developed for the advanced sodium fast reactor fuel cladding and core internals. The composition of OFRAC, which is Fe-12Cr-1Mo-03.Ti-0.3Nb + 0.3Y2O3 (wt. %), was chosen to achieve high temperature strength as well as high temperature creep performance and swelling resistance requirements that past structural materials such as austenitic stainless steels and ferritic/martensitic steels, i.e. HT-9, are not suitable as SFR cladding materials. The mechanical alloying procedures were described for producing OFRAC and the detailed characterization studies revealed uniformly distributed, high concentration of Y-Ti-O enriched oxide particles, or nanoclusters. The Nb addition that was intended to sequester interstitial C and N that are commonly introduced as contamination during ball milling of powder was found to be very effective in forming coarse size Nb carbonitrides along with excess Ti that removes these interstitial elements from the bcc-Fe matrix of OFRAC. The tensile tests on the OFRAC in the extruded condition showed excellent balance in strength and ductility from room temperature to 800°C. Both the microstructure and tensile properties were not significantly degraded after annealing the extruded OFRAC samples for 8 h at 1050°C and 1150°C. The latter temperature is ~76% of the melting point of OFRAC. These results demonstrated the stability of the microstructure and tensile properties to extremely high temperatures. The thermal creep resistance at elevated temperatures and stresses was investigated using strain rate jump testing. The results showed that OFRAC possessed high stress exponents at high stresses and temperatures of 550°C, 600°C, 700ºC and 800°C. This thermal dependence of minimum creep rate indicates that OFRAC will show quite low creep rates at higher temperatures and stresses compared to HT-9 and stainless steels. These significant improvements indicate that the NFA OFRAC alloy is a compelling and viable choice for sodium fast reactor fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Modeling and Simulation of Microstructure Evolution and Deformation in an Irradiated Environment

The ability to predict the behavior of structural components in a nuclear power plant is critical to the nuclear industry. Structural metals in the primary loop of nuclear power plants must endure challenges such as irradiation and mechanical and thermal loading, and these structural metal components must continue to function in potential transient and accident conditions throughout the operational lifetime of the power plant. This extreme operational environment changes the metal microstructure by creating additional defects. The physical interactions of dislocations with these defects govern how the metal will respond to future conditions. Therefore predicting the mechanical response of these metals requires a set of physically based and reliable models of dislocation and defect interactions. These microstructure elements include glide mobile and immobile dislocations, geometrically necessary dislocations, twinning dislocations, irradiation defects, and thermal aging defects. We present here a continuum dislocation dynamics crystal plasticity framework to capture the interaction mechanisms of these dislocations and defects, verified with a combination of benchmark problems and comparisons with experimental data for two different types of structural metals: alpha iron and nickel-based alloys. In our simulations of $\alpha$ iron we highlight the advantages of applying a Monte Carlo stochastic model of cross slip dislocation motion and show the importance of capturing the 3D nature of glide dislocation and self-interstitial atom loop radiation defect interactions. We demonstrate coupling of glide dislocations with geometrically necessary dislocations to capture the influence of lattice bending, including the sensitivity of the geometrically necessary dislocations to changes in the grain boundary angle. We further examine the interaction of glide dislocations with the twin dislocations and thermally aged defects which have been observed in a nickel-based alloy with additional models. Finally we assess the reliability of this crystal plasticity framework by comparing two dislocation glide velocity models across the range of normal operation temperatures. In successfully applying our crystal plasticity framework to multiple metals, we provide further evidence of the reliability of our approach. The results of this mechanism-based continuum dislocation dynamics crystal plasticity framework can be used to inform engineering scale models throughout the nuclear industry.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Microstructural Characterization of High Dose Irradiated HT9

The present report provides a short summary of the recent efforts to irradiate various HT9 heats and compositions to a high damage level (>500 peak dpa) and characterize the changes to the microstructure. The majority of the data reported here will focus on characterizing the “INL” HT9 heat. This particular heat was chosen due to its high volume fraction of tempered martensite (>90%) in the microstructure. The results show that very little swelling is present in INL HT9 up to doses up to 600 peak dpa. Voids are only observed in the near surface region. Some carbides are also observed to accompany the irradiation. Comparison of void swelling with published neutron irradiation shows good agreement. Additional investigations into the response of delta ferrite grains show a very different distribution of voids and significantly higher void swelling compared to the tempered martensite grains.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Microstructural Modification and Healing of Additively Manufactured Parts by Electropulsing

For many applications, the promises of additive manufacturing (AM) of rapid development cycles and fabrication of ready-to-use, geometrically-complex parts cannot be realized because of cumbersome thermal postprocessing. This postprocessing is necessary when the nonequilibrium microstructures produced by AM lead to poor material properties. This study investigated if electropulsing, the process of sending high-current-density electrical pulses through a metallic part, could be used to modify the material properties of AM parts. This process has been used to modify conventional wrought materials but has never been applied to AM materials. Two representative AM materials were examined: 316L stainless steel and A1Si10Mg. Two hours of annealing are needed to remove chemical microsegregation in AM 316L; using electropulsing, this was accomplished in 200 seconds. The ductility of AlSil0Mg parts was increased above that of the as-built material using electropulsing. This study demonstrated that electropulsing can be used to modify the microstructures of AM metals.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Using and Developing Novel Data Analytics for Digitalized Microstructures. Draft

The primary goal of this work is to develop and implement analytic methods for quantitative analysis of microstructure. These analytic methods are demonstrated on microstructures generated by the Sandia kinAic Monte Carlo code SPPARKS. This report documents progress to that end as well as incremental work towards larger aspirations of making process-structure-property connections for complex processes such as welding and additive manufacturing.

42 ENGINEERING↗
DOE OSTI2021

Thermal Conductivity Measurement of Microstructure in Irradiated Materials

Argonne National Laboratory is establishing additional capabilities for advanced techniques for accelerated experiment-based fuel material screening. A key supporting element is expanded capabilities for the characterization of irradiated samples, including establishment of the capability to measure irradiated fuel thermal conductivity at the micro-scale level. Argonne is collaborating with the University of Illinois at Urbana-Champaign (UIUC) to adopt the suspended bridge method to measure thermal conductivity of radiation-induced microstructures in nuclear materials directly. The initial phase of work to support thermal conductivity measurements of microstructure in irradiated materials at Argonne has been executed, and is summarized here. This report is intended to serve as a status update for this activity, summarizing work completed in FY20.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Microstructure and Mechanical of U-6Nb and U-8Nb in Gamma Quenched and Long-Term Aged Conditions

This report documents the mechanical response and associated microstructure in a number of recent Los Alamos studies of U-6Nb and U-8Nb alloys in the as-quenched and artificially aged conditions, to serve as a primary data reference. Quasi-static tensile and hardness data (in the form of averages) are reported along with microstructures revealed by light optical and scanning electron microscopy. Total elongation data were converted to a common standard for ease of comparison of data from different tensile specimen geometries. A variety of alternate measures of total tensile elongation confirmed the trends one would find from considering only extensometer data. The various U-Nb material pedigrees and experimental conditions are as follows. Y-12-produced U-6Nb as quenched – effect of machining damage as quenched – effect of extensometer length 200oC ages for up to 22 days Rocky Flats-produced U-6Nb 40, 65, and 90oC ages for up to 10 years 300, 350, and 400°C ages for up to 1.2 years Los Alamos-produced nonbanded U-5.6Nb and U-7.7Nb variety of ages over 100-600°C for up to 5 years We characterized the chemical banding by microprobe and Knoop microhardness surveyed over long regions in a dozen Y-12 U-6Nb specimens. As expected, the light-vs.-dark imaging response (both by light optical and scanning electron microscopy) correlated well with the local wt.% Nb. However, the hardness did not correlate well with either imaging or microprobe, and hence cannot be used to predict the local wt.% Nb.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Microstructure and Mechanical Performance of the Friction Stir Welds Made on Neutron-Irradiated Steel with Helium

This report describes new experimental results on the microstructure and mechanical performance of the friction stir welds made on neutron-irradiated steel with helium. The report focuses on helium-related issues, specifically, helium-induced degradation in the welded joint, aiming to repair irradiated components of nuclear power plants (NPPs). The friction stir welds analyzed here were produced at Oak Ridge National Laboratory previously, and limited characterization work was performed, mostly addressing the presence or absence of macroscopic cracks. The present work attempts to perform a more comprehensive study to assess microstructure conditions, grain size, plastic strain gradients, the morphology of the helium-induced damage, and deformation behavior.

42 ENGINEERING↗
DOE OSTI2019

Mechanofusion for Improved Control of Cermet TPBAR Pellet Microstructure

Mechanofusion is an alternative fabrication technique that has the potential to advance cermet TPBAR pellets by producing LiAlO 2 -Zr core-shell powders that will provide tighter control of targeted microstructural characteristics such as the presence and thickness of Zr surrounding every LiAlO 2 particle in the cermet. Significantly enhanced microstructural control will translate to improved control of material properties and characteristics that affect the in-reactor performance of cermet TPBAR pellets. It was the intention of this work to demonstrate the efficacy of mechanofusion for producing precursor powders for TPBAR pellets, however materials procurement issues prevented this objective from being fully realized. In spite of these issues, in the short amount of time during which appropriate materials were available, it was demonstrated that mechanofusion does indeed have promise for producing the desired structures although the process yet requires optimization.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructural Investigation of Precipitation Treated A709 Test Samples

This report presents the microstructural characterization using scanning transmission electron microscopy (STEM) of two creep-tested ESR1150-PT specimens, one short-term and one long-term. The results show that during creep testing, the microstructure of the A709 ESR1150-PT material continues to evolve as expected due to the synergistic effect of strain, temperature and time. Key features include dislocation forest and cell structures, precipitates on grain boundaries, precipitates randomly within grains, and precipitates associated with dislocations. In addition, the STEM study of one ESR1100-SA specimen that has been tested to rupture at a relatively low temperature (550°C) for a relatively long term (16226 h) is also presented. The results show that there is a high density of very fine nano precipitates form on dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Prediction of the influence of strain magnitude and strain path on the resulting microstructure and post-forming mechanical behavior of RA containing high formability steels

In third generation advanced high strength steels (AHSS) there is considerable potential to alter the transformation characteristics through variation of process parameters and through changes in chemical composition. This proposal aims to utilize this known phenomenon to develop microstructure-based transformation models to predict austenite stability as a function of chemical composition, strain level, deformation mode. The ultimate intent is to predict microstructure and performance during room temperature forming to produce automotive structural components with superior strength and failure resistance.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Engineering & Microstructure Scale PIE Report on EBR-II X441A Metallic Fuel Pins for the MORPH Experiment

The objectives of this project are to increase fundamental understanding of irradiation induced metallic U-Pu-Zr fuel behavior and to obtain data needed for the development of irradiation models for metallic fuels in MARMOT. The requested metallic fuel pins are from the X441A experiment irradiated in Experimental Breeder Reactor (EBR)-II and will be provided by the PIs for the duration of the experiment. The purpose of the pins irradiated in the X441A assembly was to vary Zr composition and fuel slug diameter in order to provide data for the metallic fuel performance code “LIFEMETAL” [ANL-IFR-125]. The fuels of interest to this project are metallic fuels with several Zr compositions (in wt.%): U-19Pu-6Zr, U-19Pu-10Zr, and U-19Pu-14Zr, which were irradiated to a peak burnup of approximately 11 at.%. The cladding was the same for all three fuel pins (austenitic stainless steel D9), which allows investigation of fuel-cladding interaction (FCI) phenomena. Varying Zr content in fuel pins enables investigation of the effect of Zr on fuel restructuring and fuel-cladding compatibility. Engineering and microstructure scale PIE activities will be focused on investigation of fundamental aspect of fuel performance such as species diffusion and migration, fission product behavior, and constituent redistribution. Obtained microstructural information will be used as the basis for the development of MARMOT models of U-Pu-Zr fuel performance at the mesoscale. Experimental data obtained through this proposal will be used to provide this fundamental understanding which will serve as the foundation of the development of radiation models for U-Pu-Zr in MARMOT. It will also provide a starting point for the design of new experiments to provide data for the validation of these MARMOT models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Microstructure and in-service degradation of baffle former bolts – in-core components of light water reactors

This report presents a microstructure evaluation of the baffle former bolt, a commercial pressurized water reactor component. Using SEM, EDS, and EBSD analysis the material was identified as AISI 316 steel with an annealed austenite microstructure. Findings include abnormal grain growth in a fraction of the grains and the absence of retained ferrite. Notable observations include an overlapping of pre- and post-irradiation deformation, including defect-free channel formation as an active deformation mechanism in the near-surface layer of BFB. Specimen surfaces, exposed to high-temperature, high-pressure water, exhibited signs of in-service corrosion degradation. EBSD and EDS analyses highlighted intergranular corrosion, possible grain boundary oxidation with depths less than 3 µm, and, unexpectedly, short cracks filled with Cr-rich oxides measuring around 5-6 µm. The presence of defect-free channels from EBSD data suggests episodes of high mechanical stresses during service. The specimens provide a potentially unique insight into the in-service degradation of NPP components, strain localization, and crack initiation processes. An intact corrosion layer reveals a complex component's in-service history. Further research is underway for a detailed analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Investigation on laser absorption and x-ray radiation in microstructured titanium targets heated by short-pulse relativistic laser pulses

The enhancement effect of a microstructured surface on laser absorption and characteristic Kα emission has been investigated by measuring K-shell emission from titanium (Ti) targets irradiated with high-intensity (~10 20 W cm -2 ), subpicosecond (500 fs) laser pulses. The experimental results indicate a modest enhancement (1.6×) of Kα emission from microstructured targets compared to flat foils, but with a similar intensity and profile of Heα and Li-like satellites. Particle-in-cell (PIC) simulations are implemented to further understand the underlying physical processes in the laser interaction with both targets, interpreting the mechanisms responsible for the Kα enhancement. The reasons for the lower-than-expected enhancement of Kα emission are discussed. The rapid heating of the bulk plasma might result in the premature shutdown of Kα emission before the thermalization of hot electrons or even the end of laser pulses, suggesting that the use of Kα emission as a diagnostic of the hot-electron yield or relaxation could lead to a misinterpretation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructural Characterizations of Two High Fluence Baffle-Former Bolts Retrieved from a Westinghouse Two-loop Downflow Type PWR

As one of the pressurized water reactor (PWR) internal components, baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts. In support of evaluating long-term operational performance of materials used in core internal components, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Office of Nuclear Energy, Light Water Reactor Sustainability (LWRS) Program, Materials Research Pathway (MRP), has harvested two high fluence BFBs from a commercial Westinghouse two-loop downflow type PWR. The two bolts of interest, i.e. bolts # 4412 and 4416, were withdrawn from service in 2011 as part of a preventative replacement plan. No identification of cracking or potential damage was found for these bolts during their removal in 2011. However, the bolts required a lower torque for removal from the baffle structure than the original torque specified during installation. Irradiation damage levels in the bolts range from 15 to 41 displacements per atom. The goal of this project is to perform detailed microstructural and mechanical property characterization of BFBs following in-service exposures. The information from these bolts will be integral to the LWRS program initiatives in evaluating end of life microstructure and properties. Furthermore, valuable data will be obtained that can be incorporated into model predictions of long term irradiation behavior and compared to results obtained from experimental reactor conditions

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Report on the Integration of Experimental and Modeling Data for Initial Equivalence Study of Microstructural Evolution in Irradiated LPBF 316SS

Advanced materials and manufacturing technologies are poised to improve the safety and design characteristics of nuclear technologies and meet US energy, environmental, and economic needs. In particular, metal additive manufacturing (AM) provides an opportunity to produce novel materials and component geometries, but their use is not without hurdles arising from the inherent microstructure variability that can result from the layer-by-layer build approach. Given the greater possible microstructure variability in AM materials—and the dearth of materials test reactors—it is impractical to rely solely on neutron irradiation studies to produce data for materials qualification for every possibility. This work within the Advanced Materials and Manufacturing Technologies (AMMT) Environmental Effects technical area contributes to the rapid qualification framework by developing a science-driven framework for the accelerated qualification of materials for nuclear environments. A key product of the Environmental Effects technical area of the AMMT program is the Licensing Approach with Ions and Neutrons (LAIN). This approach recognizes that whether using existing materials in new environments, newly developed materials tailored for these environments, or new manufacturing methods, the traditional decades-long approach for materials qualification does not facilitate rapid deployment. In FY 2023, the AMMT program presented a conceptual framework of specific steps to fulfill several technical challenges associated with qualifying materials for performance in radiation environments on an accelerated time frame informed by the state of the art in materials science and a review of the current regulatory landscape. The objective of this section of the Environmental Effects technical area is to critically evaluate and refine the proposed qualification framework presented under AMMT by integrating the research results of the neutron irradiations, the ion irradiations, and modeling efforts. These ongoing efforts span across Argonne National Laboratory (ANL), Idaho National Laboratory (INL), and Oak Ridge National Laboratory (ORNL) and are closely coordinated.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Microstructural Engineering and Accelerated Test Method Development to Achieve Low Cost, High Performance Solutions for Hydrogen Storage and Delivery

This project made advancements in developing lower cost steel alloys with novel microstructural design for use in hydrogen refueling infrastructure such as storage, compressors, and dispensing components, and utilizing accelerated test methods to efficiently evaluate variations in alloy and microstructure design. The project specifically sought to design alloys with lower nickel contents to reduce alloy cost, which was accomplished through substituting manganese for nickel along with other alloy additions to control deformation characteristics known to be important for hydrogen embrittlement resistance. Through Mn substitutions for nickel, austenitic and duplex austenite-ferrite steels were successfully developed with lower cost than currently available commercial stainless steel products that are employed for hydrogen refueling infrastructure. The steels were processed to achieve comparable strength and toughness in hydrogen environments as the commercially available steels containing high Ni contents. To evaluate mechanical performance in hydrogen, a testing methodology was employed to compare ubiquitous laboratory testing using electrochemical hydrogen charging in a liquid electrolyte to less accessible high pressure gaseous testing. While the application of these steels is in high pressure gaseous environments, the electrochemical hydrogen charging tests produced comparable results. Additionally, the Los Alamos Neutron Scatting Center enabled characterization of deformation mechanisms of the steel alloys in the presence of hydrogen, which has been associated with steel alloy characteristics associated with hydrogen embrittlement. Finally, a fracture mechanics based test bed model was developed to predict the influence of hydrogen gas pressure and fatigue conditions on fatigue lifetimes of pressure vessel steels. Together, these developments can be employed to enable lower cost hydrogen fueling infrastructure and more reliable prediction of steel alloy components in hydrogen service conditions. In particular, the newly alloys have the potential to replace stainless steels, having demonstrated comparable performance at substantially reduced cost.

08 HYDROGEN↗