Results for “as-fabricated” — Page 11

NASA NTRS2018

Impact of Powder Variability on the Microstructure and Mechanical Behavior of Selective Laser Melted (SLM) Alloy 718

Nickel Alloy 718 has found wide use in high-temperature aerospace applications (up to 700°C), due to a combination of good mechanical properties, environmental resistance, and workability at moderate cost. These properties have led Alloy 718 to early adoption by selective laser melting (SLM) additive manufacturing. This presentation will share results from a comprehensive industry survey of 718 powder feedstock for SLM. The survey focuses on the impact of variability in powder characteristics on build quality, microstructure, tensile and high cycle fatigue in the as-fabricated and machined surface conditions. The investigation includes sixteen powder lots from eight suppliers that are produced by gas atomization or rotary atomization in argon or nitrogen. This research was previously presented and published at the 9th International Symposium on Superalloy 718 & Derivatives and was funded by NASA Space Launch System Liquid Engine Office with the Human Exploration Operations Mission Directorate.

Sudbrack, Chantal↗

NASA NTRS2019

Electrical and Thermal Characterization of 3D Printed Thermoplastic Parts with Embedded Wires for High Current-Carrying Applications

Fabrication of parts exhibiting multi-functionality has recently been complemented by hybrid polymer extrusion additive manufacturing in combination with wire embedding technology. While much mechanical characterization has been performed on parts produced with fused deposition modeling, limited characterization has been performed when combined electrical and thermal loads are applied to 3D printed multi-material parts. As such, this work describes the design, fabrication, and testing of 3D printed thermoplastic coupons containing embedded copper wires that carried current. An automated fabrication process was used employing a hybrid additive manufacturing machine that dispensed polycarbonate thermoplastic and embedded bare copper wires. Testing included AC and DC hipot testing as well as thermal testing on as-fabricated and heat treated coupons to determine the effect of porosity in the substrate. The heat-treated parts contained reduced amounts of porosity, as corroborated through scanning electron microscopy, which led to 50 % increased breakdown strength and 30 to 40 % increased heat dissipation capabilities. The results of this research are describing a set of design protocol that can be used as a guideline for 3D printed embedded electronics to predict the electrical and thermal behavior.

Hipot testing↗

NASA NTRSDate not supplied

TPSAS-NF1676L-34417-DND

A boron nitride nanotube (BNNT), an electrically-insulating counterpart of a carbon nanotube (CNT), is a great nanofiller candidate for ferroelectric polymeric nanocomposites due to BNNT?s exceptional properties in piezoelectricity, thermal stability, mechanical strength, and radiation shielding capability. Fast and high-yield polar crystallization of polyvinylidene fluoride (PVDF) was accomplished by incorporating BNNTs as a nucleating agent. As-fabricated BNNT-PVDF nanocomposites were characterized by Fourier-transform infrared spectroscopy (FTIR), wide-angle X-ray diffraction (XRD), and differential scanning calorimetry (DSC) to identify and quantify chain conformations. Besides, piezoelectric constants, d33 of BNNT-PVDF nanocomposites were measured and compared with that of a mechanically drawn beta phase polar PVDF. The polar transformation due to BNNT incorporation has a great advantage over conventional beta or gamma transformation methods that require (1) mechanical drawing often resulting in defects or (2) high-temperature annealing for an extended time. In addition to the fast polar crystallization, BNNTs had a role of reinforcement of the polymer matrix. The improved Young?s modulus and electromechanical coupling coefficient of BNNT-PVDF nanocomposites indicated potential applications in energy harvesting under harsh environments such as large deformation, wide temperature cycles, and high radiation.

Dongwon Lee↗

NASA NTRSDate not supplied

TPSAS-NF1676L-17800-DND

Currently, there are two national challenge problems that guide much of the research in durability and damage tolerance at NASA Langley. The first, Airframe Digital Twin, is a concept that combines as-built vehicle components, as-experienced loads and environments, and other vehicle-specific characteristics to enable ultrahigh fidelity modeling of aircraft and spacecraft throughout their service lives. The second, Materials Genome Initiative, is an analog to the Human Genome Project, and is intended to improve the rate at which materials scientists can discover, understand fundamental physics, and improve material systems. This presentation will highlight several research projects ongoing at NASA Langley that are in support of the above challenge problems. Two of those topics will be the subject of detailed discussion. First, investigations of microstructurally-small fatigue cracking (MSFC) in Al-2Cu and Al-4Cu, fabricated in-house, will be presented. Single- and oligo-crystals of Al-Cu specimens were loaded in uniaxial fatigue, while high-resolution in-situ measurements of deformation were made using image correlation (IC) in a scanning-electron microscope (SEM). The Al-Cu specimens were then replicated as crystal plasticity finite element models (CPFEM), where evolution of slip localization near grain boundaries was computed. Comparison among experiment and CPFEM is made. In addition, XRay diffraction measurements of the as-fabricated specimens were carried out, where direct measurements of the embedded copper precipitates were made, and their influence on growing MSFCs were directly observed. The second main topic will illustrate ongoing work in the area of so-called damage-sensing particles. In this work, shape-memory alloys are embedded in an aluminum alloy matrix. Upon the propagation of a fatigue crack, these particles undergo a strain-induced phase transformation which is detected using an acoustic sensor, providing real-time information on propagating cracks. Experiments and simulations regarding the development of this system will also be detailed.

Jacob Hochhalter↗

NASA NTRS2021

Effect of Temperature History on Thermal Properties of Additively Manufactured C-18150 Alloy Samples

The effects of thermal history on the thermal properties of C-18150 (Cu-1.5Cr-0.5Zr, wt.%) parts made by the laser powder-bed-fusion (L-PBF) additive manufacturing (AM) process with three different fabrication orientations (horizontal, angled, and vertical to the build direction) were investigated. A significant difference in thermal properties between as-fabricated and fully heat-treated L-PBF C-18150 samples was observed, and the mechanisms of such behaviors are discussed.

C-18150↗

NASA NTRSDate not supplied

Hot Structure Concepts and Materials for Space Exploration Project

Hot structures are multifunctional material systems that integrate the functions of thermal protection systems (TPS) and primary structure to increase mass efficiency in aerospace vehicles. The goal of this project was to improve our capabilities for developing lightweight, reusable, and reliable hot structures concepts to enable advanced space exploration missions. Specifically, current hot structure carbon/carbon (C/C) composite material systems have low interlaminar strength, which can lead to material failures at low load levels, and large material property variability in as-fabricated parts. Two parallel paths were followed to address these shortcomings. First, analysis models were developed to identify the conditions under which excessive interlaminar stresses occur and to evaluate the potential of stitching the C/C laminae to improve the interlaminar thermal-structural response. This effort included the development of high-fidelity, fracture-based thermal-structural finite element analysis models to evaluate critical features needed to produce efficient hot structure concepts. The second path focused on developing atomistic and mesoscale simulation tools to model the fabrication process in an effort to reduce the time and cost of production while improving the quality of the resulting parts. Aspects of the process currently being investigated include the infusion of the matrix precursor into the fiber preform, the cure of the precursor molecules into their polymeric form, and the carbonization and graphitization of the polymer to form the carbon matrix material. Results of the 6-month study are presented herein.

thermal↗

NASA NTRSDate not supplied

Effect of Heat Treatment on the Microstructure and Mechanical Properties of Monel K500 Alloy Fabricated Via L-PBF and LP-DED

This study examines and compares the effect of different heat treatments (HT) on the microstructure and mechanical properties of Monel K500 fabricated using laser powder bed fusion (L-PBF) and laser powder directed energy deposition (LP-DED) technologies. The as-fabricated Monel K500 specimens exhibited dendritic microstructure and elemental micro-segregation due to high cooling rates induced during the fabrication process. The applicability of HT proposed in the literature for wrought Monel K500 was investigated for L-PBF and LP-DED using four different HT procedures involving hot isostatic pressing (HIP), solution annealing (SA), and aging. The mechanical properties of test specimens were evaluated using uniaxial tensile testing at room temperature. The microstructural evolution of test specimens during HT was analyzed using a scanning electron microscope. For all HT conditions investigated, L-PBF Monel K500 specimens consistently displayed higher strength and lower ductility compared to the LP-DED counterparts. The HT procedure involving HIP at 1160°C for 3hr at 100 MPa, SA at 1100°C for 15 min, and three step aging at 610°C for 16 hr, 540°C for 6 hr, and 480°C for 8 hr resulted in highest strength for both L-PBF and LP-DED fabricated Monel K500.

Additive manufacturing↗

NASA NTRS2023

A Summary of Structural Test and Simulation Results for Hawker 4000 Composite Fuselage Hardware

NASA Langley Research Center (LaRC) was able to obtain a partially fabricated Hawker Beechcraft Model 4000 (Hawker 4000) aft aircraft section to use for study. The Hawker 4000 is a medium sized business jet (14 Code of Federal Regulations Part 25 – Transport Category Aircraft) with seating for 8 to 14 people and fabricated out of carbon composite. The Hawker 4000 was designed in the early 2000s to compete with Gulfstream and Bombardier business jet units. The Hawker Beechcraft business unit declared bankruptcy in 2012, and abruptly halted production of the 4000 series aircraft. As a result, only 73 full production aircraft were built. This all-composite aft fuselage section was an ideal candidate to utilize for investigations of composite structures undergoing dynamic loading in order to satisfy NASA Revolutionary Vertical Lift Technology (RVLT) project objectives. Thus, a team at NASA LaRC embarked on a test and simulation program in late 2021 in order to generate test data of composite structures undergoing dynamic loading conditions. As a part of these investigations, methods including reverse engineering the geometry, conducting as-fabricated materials testing and using non-destructive evaluation (NDE) techniques such as ultrasonic through transmission were utilized in order to develop digital twin finite element models for use in computational simulations.

dynamic composite testing↗

DOE OSTI2021

Advanced Fuels Campaign Execution Plan

The Advanced Fuels Campaign (AFC) Execution Plan outlines the strategy, mission, scope, near-term and long-term goals, structure, and organization associated with nuclear fuels and materials research, development, and demonstration activities within the Department of Energy’s (DOE) Nuclear Fuel Cycle and Supply Chain (NFCSC) program. NFCSC has been given responsibility to identify and mature advanced fuel technologies for the DOE using a science-based approach, focused on developing a fundamental understanding of nuclear fuels and materials to drive development of integrated nuclear fuel and materials technology. This science-based approach combines theory, experiments, and multiscale modeling and simulation to achieve a predictive understanding of relevant behaviors ranging from fuel fabrication processes (and their resulting fuel microstructures) through fuel/cladding performance under irradiation (in contrast to more empirical, observation-based approaches frequently used in fuel performance modeling and fuel qualification). The traditional scope of AFC includes the evaluation and development of multiple fuel forms to support two fuel cycle options: once-through and full recycle. The word “fuel” is used generically to include conventional fuels, transmutation targets, and any associated cladding or duct materials. The once-through fuel cycle addresses advanced light water reactor fuels with enhanced performance, extended burnup, and reduced waste generation. In fiscal year (FY) 2012, AFC’s scope expanded to include research, development, and demonstration (RD&D) for light water reactor (LWR) fuels with enhanced accident tolerance. Fuel fabrication activities include the development of innovative methods to enhance process efficiencies, reduce waste, and improve control over as-fabricated fuel microstructural properties to achieve desired in-reactor performance. Using modern modeling and simulation approaches, the objective is to predict fresh fuel properties given the feedstock characteristics and fabrication process parameters. The performance-related activities include small-scale, in-reactor, and out-of-reactor phenomenological testing (distinct from, but synergistic with, integral prototypic testing) and extensive, quantitative characterization (focusing on characterization of fuel and cladding materials at the scale of microstructure) both before and after testing. Larger-scale, prototypic experiments are conducted in concert with phenomenological testing to drive a Fuel Development and Qualification program, incorporating a fundamental understanding of fuel behavior performance characteristics. Then, using the tools developed under the productive science-based approach, fuels will be optimized to meet specific performance requirements, thereby minimizing the need to repeatedly perform large-scale, integral experiments over a wide parametric range as a means of experimental exploration. Two significant initiatives are underway within AFC. First, a gap analysis completed in early FY 2019 identified critical irradiation testing needs that are lacking within the national light water reactor (LWR) fuels testbed since the shutdown of the Halden Reactor in 2018. The identified gaps are for instrumented, prototypic testing of LWR fuels, especially under boiling water reactor conditions, ramp conditions, and conditions leading to fuel failure; these needs exist for supporting current LWR fuels and their possible extension to higher burnups, but are especially urgent relative to near-term development and qualification of accident-tolerant fuels. Recommendations that resulted from the Halden Gap Analysis focused on enhancements at Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) to fill gaps in testing capabilities relative to these needs. Second, a concerted effort to develop and demonstrate a systematic approach to accelerating the development, testing, and qualification of new fuel systems has been initiated. This is highlighted by a test strategy that combines the considerable advances in multiscale, mechanistic fuel modeling of recent years with a MiniFuel separate effects test program in the High Flux Isotope Reactor (HFIR) and a Fission Accelerated Steady-state Testing (FAST) semi-integral accelerated test program in ATR. This approach is being tested/demonstrated using the metallic fuel system, but if successful it is expected to be extensible to multiple fuel types and diverse applications. This document includes an overview of the NFCSC program, a definition of science-based development of nuclear fuels, near-term goals for Advanced LWR fuels (ALFs), and longer-term goals for Advanced Reactor Fuels (ARFs) RD&D. This includes the activities that will be conducted to achieve success toward the grand challenge, as well as the goals and milestones to be achieved over the next few decades of research and development. Long-term goals are based on the DOE Office of Nuclear Energ

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Graphite Sessions 1-3

International Research & Development Programs Who is doing what in graphite throughout the world Nuclear graphite components Why is graphite so difficult to develop code rules? Unique features of nuclear grade graphite No nuclear graphite standard Brittle material in a pressure vessel code? General graphite behavior and degradation As-fabricated material properties Irradiation behavior Oxidation behavior Molten salt issues

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Destructive Examination of a FeCrAl-UO 2 Irradiation Test

Several destructive postirradiation examinations were performed on an irradiation specimen that coupled an early iron-chromium-aluminum (FeCrAl) candidate alloy cladding with UO2 pellets. This irradiation test was designed to investigate the early life performance and compatibility of the FeCrAl-UO 2 cladding-fuel system under prototypic light water reactor neutronic conditions. Additionally, these tests were expected to provide neutron irradiated samples for severe accident testing of this fuel system. The irradiation studied in this work was part of the ATF-1 series of drop-in style irradiations performed in the Idaho National Laboratory Advanced Test Reactor. The rodlet studied in this work is one of three similar rodlets irradiated in ATF-1 that had approximately 7.6 cm of cladding machined from a wrought FeCrAl alloy and fueled with a 6.1 cm stack of UO 2 pellets. The outer diameter of the cladding was ~0.94 cm and the inner diameter was ~0.83 cm. After irradiation and non-destructive examination in Idaho this rodlet was shipped to the Oak Ridge National Laboratory hot-cells for further examination. The irradiated rodlet was sectioned into several samples for microstructural, micromechanical, and severe accident testing. During sectioning, it was noted that the fuel was not firmly bonded to the cladding and could be readily removed from small cladding slices. Microstructural characterization of fuel cross sections also revealed no significant interaction between the fuel and the cladding. Samples were also prepared for microhardness testing. To prepare for high temperature oxidation testing, the fuel was dissolved from segments of the cladding. High temperature oxidation testing of cladding segments was performed at 1200°C and 1300°C in a steam environment. Comparisons between the oxidation of this FeCrAl alloy in its neutron irradiated state, as-fabricated state, and the oxidation of Zircaloy-2 are made. The oxidation testing will be followed by ring compression testing to evaluate ductility. The microstructure of the samples after oxidation and ring compression testing will also be analyzed.

Harp, Jason↗

DOE OSTI2022

AGR-5/6/7 Final Release-to-Birth Ratio Data Analysis

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel for use in the high-temperature gas cooled reactor (HTGR). AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product (FP) transport models and codes. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. Release-to-birth ratios (R/B) for fission-gas isotopes released from each capsule are calculated from release rates, measured by germanium detectors in the Fission Product Monitoring System (FPMS) installed downstream from each capsule, and birth rates calculated using numerical models of FG generation. The R/Bs are a critical measure of the ability of the fuel kernel, the particle coating layers, and the compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, and the impact of initially defective particles and/or particle-coating failures that occur during irradiation. For fission-gas isotopes, particle failure is defined as failure of all coating layers, allowing gaseous fission atoms to escape from a particle. During the first five cycles (162B ? 165A), R/Bs were stable in the 10-8?10-6 range, and no in-pile particle failures were observed, based on the gross gamma counts. The maximum R/B value of around 2 ? 10-6 for Kr-85m resulted from the presence of as-fabricated exposed kernels (based on the high exposed kernel fraction), the dispersed uranium, and high fuel particle temperatures in Capsule 1. Comparison of capsule-measured R/Bs from these early cycles to predictions using the previously developed AGR R/B model demonstrated FG release from the AGR-5/6/7 TRISO fuel was comparable to that of previous experiments. In addition, the Kr-85m R/B per-exposed-kernel values are comparable to R/B values obtained in AGR-3/4 irradiation experiment and four irradiation experiments performed during 1980s: (1) HRB-17/18, (2) COMEDIE-BD1, (3) HFR B1, and (4) HRB-21. In contrast, all measured R/B values are lower than predictions by the commonly used Richards and German models, which are intentionally conservative. A large number of in-pile particle failures occurred in Capsule 1 by the end of Cycle 166A. During the final four cycles (166A ? 168A), apparent damage to the Capsule 1 gas line appeared to cause FG leakage from that capsule into the other four capsules, resulting in an increase in fission gas (FG) detected in the effluent for all capsules. Isolation of the Capsule 1 gas line during the last three cycles also prevented measurement of its FG release. Thus, R/Bs in all capsules after Cycle 166A are highly uncertain because of undefined amount of leakage from Capsule 1, especially for long-lived isotopes. A few hundred in-pile particle failures were estimated for Capsule 1 before the end of Cycle 166A, but the total number of failures is unknown due to the lack of FG release data in the later cycles. Based primarily on evidence from the gross gamma counts during Cycle 168A, approximately 15 particles failed in Capsule 3 and four particles failed in Capsule 2. In-pile failures in Capsule 3 were anticipated because this capsule was designed to operate beyond the HTGR normal operating temperature range. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR model, developed in INL/EXT-14-32970, for R/B of the short-lived isotopes (Kr-89 and Xe-137) with minimal leakage from Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Graphite Material Behavior: NRC Graphite Behavior Model

Introduction to graphite Fabrication of graphite Unique features of nuclear grade graphite Providing the properties As-fabricated material properties of interest Thermal diff, thermal stability, mechanical strength, etc. General graphite behavior and degradation Microstructure: key to behavior Crystallographic structure Basal plane (covalent) bonding Porosity and pore microstructure Degradation Irradiation behavior Oxidation behavior Molten salt issues

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Corrosion testing needs and considerations for additively manufactured materials in nuclear reactors

Metal additive manufacturing holds significant promise as an enabling technology for the 21st century nuclear energy industry. Metal additive manufacturing (MAM) can allow the fabrication of novel materials and innovative component designs that are not achievable through conventional manufacturing. Due to its very different fabrication methods, as-fabricated MAM components are characteristically different from conventionally manufactured components. MAM materials exhibit very different microstructures from conventional cast or wrought materials. For example, austenitic stainless steels fabricated by laser powder bed fusion exhibit columnar grain structures, dislocation cell structures, and melt pool fingerprints. In addition, MAM fabrication may result in defects such as porosity, incomplete processing of the feedstock (e.g., lack of fusion in melt-based methods), and oxide inclusions. Heat treatments may further evolve the microstructure, microsegregation, and stresses within the component. Furthermore, MAM components have a rough surface with feature sizes on the order of the feedstock material, as opposed to smooth surfaces resulting from conventional machining and forming operations. As a result, the corrosion behavior of MAM components will likely be significantly different from that of conventionally formed components. Corrosive environments for structural materials within advanced reactor environments include molten fluoride and chloride salts, liquid sodium and lead-bismuth, and high-temperature helium. The Advanced Materials and Manufacturing Technologies program within the Department of Nuclear Energy in the United States Department of Energy is assessing the unique concerns of MAM component corrosion and testing methodologies in advanced nuclear reactor environments. We discuss these concerns and testing strategies in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Advanced Fuels Campaign Execution Plan

The Advanced Fuels Campaign (AFC) Execution Plan outlines the strategy, mission, scope, near-term and long-term goals, structure, and organization associated with nuclear fuels and materials research, development, and demonstration activities within the Department of Energy’s (DOE) Nuclear Fuel Cycle and Supply Chain (NFCSC) program. NFCSC has been given responsibility to identify and mature advanced fuel technologies for the DOE using a science-based approach, focused on developing a fundamental understanding of nuclear fuels and materials to drive development of integrated nuclear fuel and materials technology. This science-based approach combines theory, experiments, and multiscale modeling and simulation to achieve a predictive understanding of relevant behaviors ranging from fuel fabrication processes (and their resulting fuel microstructures) through fuel/cladding performance under irradiation (in contrast to more empirical, observation-based approaches frequently used in fuel performance modeling and fuel qualification). The traditional scope of AFC includes the evaluation and development of multiple fuel forms to support two fuel cycle options: once-through and full recycle. The word “fuel” is used generically to include conventional fuels, transmutation targets, and any associated cladding or duct materials. The once-through fuel cycle addresses advanced light water reactor fuels with enhanced performance, extended burnup, and reduced waste generation. In fiscal year (FY) 2012, AFC’s scope expanded to include research, development, and demonstration (RD&D) for light water reactor (LWR) fuels with enhanced accident tolerance. Fuel fabrication activities include the development of innovative methods to enhance process efficiencies, reduce waste, and improve control over as-fabricated fuel microstructural properties to achieve desired in-reactor performance. Using modern modeling and simulation approaches, the objective is to predict fresh fuel properties given the feedstock characteristics and fabrication process parameters. The performance-related activities include small-scale, in-reactor, and out-of-reactor phenomenological testing (distinct from, but synergistic with, integral prototypic testing) and extensive, quantitative characterization (focusing on characterization of fuel and cladding materials at the scale of microstructure) both before and after testing. Larger-scale, prototypic experiments are conducted in concert with phenomenological testing to drive a Fuel Development and Qualification program, incorporating a fundamental understanding of fuel behavior performance characteristics. Then, using the tools developed under the productive science-based approach, fuels will be optimized to meet specific performance requirements, thereby minimizing the need to repeatedly perform large-scale, integral experiments over a wide parametric range as a means of experimental exploration. Two significant initiatives are underway within AFC. First, a gap analysis completed in early FY 2019 identified critical irradiation testing needs that are lacking within the national light water reactor (LWR) fuels testbed since the shutdown of the Halden Reactor in 2018. The identified gaps are for instrumented, prototypic testing of LWR fuels, especially under boiling water reactor conditions, ramp conditions, and conditions leading to fuel failure; these needs exist for supporting current LWR fuels and their possible extension to higher burnups, but are especially urgent relative to near-term development and qualification of accident-tolerant fuels. Recommendations that resulted from the Halden Gap Analysis focused on enhancements at Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) to fill gaps in testing capabilities relative to these needs. Second, a concerted effort to develop and demonstrate a systematic approach to accelerating the development, testing, and qualification of new fuel systems has been initiated. This is highlighted by a test strategy that combines the considerable advances in multiscale, mechanistic fuel modeling of recent years with a MiniFuel separate effects test program in the High Flux Isotope Reactor (HFIR) and a Fission Accelerated Steady-state Testing (FAST) semi-integral accelerated test program in ATR. This approach is being tested/demonstrated using the metallic fuel system, but if successful it is expected to be extensible to multiple fuel types and diverse applications. This document includes an overview of the NFCSC program, a definition of science-based development of nuclear fuels, near-term goals for Advanced LWR fuels (ALFs), and longer-term goals for Advanced Reactor Fuels (ARFs) RD&D. This includes the activities that will be conducted to achieve success toward the grand challenge, as well as the goals and milestones to be achieved over the next few decades of research and development. Long-term goals are based on the DOE Office of Nuclear Energ

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Micro-tensile Properties of Fueled Irradiated AGR-2 TRISO-coated Particle Buffer, IPyC, and SiC Interlayer Regions

Tristructural isotropic (TRISO) coated nuclear fuel particles are proving to be a versatile fuel form for new reactor designs. Understanding the bounding strength and failure mode of each coating interface is important to both fuel quality evaluation and failure prediction. A mechanism of key significance is failure of the silicon carbide (SiC) layer to retain fission products due to incomplete tearing of the buffer layer. This is a two-step mechanism involving both mechanical failure in the buffer and inner pyrolytic carbon (IPyC) layers and degradation of the SiC layer through palladium silicides at the IPyC-SiC interface. However, the mechanical properties of TRISO particle coating layers have yet to be fully characterized due to the small dimension of TRISO fuel particles and high radioactivity. To investigate this mechanism, in situ micro-tensile properties of the buffer, IPyC, SiC, buffer-IPyC, and IPyC-SiC interlayer regions of fueled TRISO particles have been tested at both as-fabricated and irradiated conditions. Determination of the mechanical properties of these TRISO particle regions will lead to a better understanding of the SiC layer failure mechanism and enable progress towards TRISO fuel qualification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Mapping the Microstructure During the Historic U-Mo Monolithic Fuel Foil Fabrication Process

Since 2004, there has been extensive effort towards the development of a uranium molybdenum monolithic fuel system to convert high performance research and test reactors. The RERTR-6 experiment was the first to attempt a monolithic fuel instead of a dispersed fuel form. The fabrication methods evolved overtime and provided the basis for current fabrication methods. The various steps in the process inevitably tailor the fuel alloy microstructure which is known to influence irradiation behavior. This document aims to present and discuss the fabrication evolution that transpired through a recounting of historical data from the various fabrication campaigns. By overlaying this data with basic science studies on the U-Mo that explored transformation kinetics, it is possible to estimate a measure of the impact heat treatments have on the final as-fabricated microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

DEVELOPMENT OF INEXPENSIVE HIGH TEMPERATURE NITI-BASED SHAPE MEMORY ALLOYS FOR POWDER BED ADDITIVE MANUFACTURING

NiTi and NiTi-based Shape Memory Alloys (SMA) exhibit a reversible solid-state phase transformation from martensite to austenite driven by thermal energy. High temperature (Mf>100°C) SMAs are martensite at room temperature and can be fabricated into solid-state actuators that return to a pre-programmed shape against a designed load after heating to transformation threshold. Reactive as-fabricated additively manufactured parts (4-D printing) is the current state of the art in manufacturing of SMAs but requires compositions compliant to rapid solidification. Existing actuator designs are developed from commercially available, highly investigated material compositions. However, existing high temperature high performance (high actuation strain, low thermal hysteresis) shape memory alloys contain significant (>10% at.) portions of high-cost Platinum Group Metals (PGMs). It is of significant scientific interest to investigate material compositions that are peer performing or superior to PGMs whose constituent elements represent a significant cost savings. Shape memory alloy properties vary significantly with small (0.1% at.) compositional changes making robust investigative sample sets very large. Computational material design can be deployed to shrink the compositional space of possible alloy combinations and reduce the experimental load in material discovery. Investigating shape memory effect (SME) and validating process additive process parameters for a single novel composition is cost intensive in both time and consumed materials. Additionally, sub-optimal processing, oxygen, or solidification rate sensitivity could render additively manufacturing specimens without micro, macro cracks, or significant chemical variance impossible. Unfortunately, such failure susceptibility cannot be simulated. Therefore, a research pathway to validate novel shape memory alloy compositions for powder bed fusion additive manufacturing without the need for powdered feedstock is also proposed. This research investigates novel high temperature shape memory alloys for actuators without platinum group alloying elements to discover one that could be commercially viable as an additive manufacturing feedstock.

Sundermann, Tayler↗