GCR: Development of Improved Alloy 800H Weldment
Alloy 800H welding filler metal project presentation for the 2023 Advanced Reactor Technology (ART) program review.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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Perspectives on Quenching and Tempering 4340 Steel
Steels are ubiquitous due to their affordability and the landscape of useful properties that can be generated for engineering applications. But to further expand the performance envelope, one must be able to understand and control microstructure development by alloying and processing. In this paper we use multiscale, advanced characterization to better understand the structural and chemical evolution of AISI 4340 steel after quenching and tempering (Q&T), including the role of quench rate and short-time, isothermal tempering below 573 K (300 °C), with an emphasis on carbide formation. We compare the microstructure and/or property changes produced by conventional tempering to those produced by higher temperature, short-time “rapid” tempering. We underscore that no single characterization technique can fully capture the subtle microstructure changes like carbon redistribution, transition carbide and/or cementite formation, and retained austenite decomposition that occur during Q&T. Only the use of multiple techniques begins to unravel these complexities. After controlled fast or slow quenching, η transition carbides clearly exist in the microstructure, likely associated with autotempering of this high martensite start temperature (Ms) steel. Isothermal tempering below 598 K (325 °C) results in the relief of carbon supersaturation in the martensite, primarily by the formation of η transition carbides that exhibit a range of carbon levels, seemingly without substitutional element partitioning between the carbide and matrix phases. Hägg transition carbide is present between 300 °C and 325 °C. After conventional tempering at or above 598 K (325 °C) for 2 h, cementite is predominant, but small amounts of cementite are also present in other conditions, even after quenching. Previous work has indicated that silicon (Si) and substitutional elements partition between the cementite, which initially forms under paraequilibrium conditions, and the matrix. Phosphorous (P) may also be preferentially located at cementite/matrix interfaces after high temperature tempering. Slower quench rates result in greater amounts of retained austenite compared to those after fast quenching, which we attribute to increased austenite stability resulting from “autopartitioning”. Rapid, high temperature tempering is also found to diminish tempered martensite embrittlement (TME) believed to be associated with the extent of austenite decomposition, resulting in mechanical properties not attainable by conventional tempering, which may have important implications with respect to industrial heat treatment processes like induction tempering. Controlling the amount and stability of retained austenite is not only relevant to the properties of Q&T steels, but also next-generation advanced high strength steels (AHSS) with austenite/martensite mixtures.
36 MATERIALS SCIENCE↗
Mechanisms for high creep resistance in alumina forming austenitic (AFA) alloys
Castable alumina forming austenitic (AFA) alloys have demonstrated superior creep life and oxidation resistance at temperatures exceeding 800⁰C. Despite the success in the applicability of these alloys in extreme environments, there is a limited understanding of the deformation modes and the influence of each alloying element guiding the alloy design strategies that could further enhance the creep strength of these AFA alloys, particularly at temperatures at and above 900⁰C. In this study, we reveal the mechanism underpinning the superior creep performance of castable AFA alloys that involves suppressing primary carbide formation through minor compositional modification. This approach results in a three-fold increase in creep strength at 900⁰C and 50 MPa. Here, by employing integrated characterization techniques, we analyzed the microstructures of two AFA alloys, both before and after the creep process. We discovered that the suppression of primary carbides permits the in-situ clustering of now-available interstitial elements such as C, Si, and O during high-temperature creep. This improved solid solution strengthening and reduced stacking fault energy of the alloy. Moreover, it also enabled controlled secondary carbide formation during testing, further improving the creep resistance. These findings underline the important interplay between alloy composition, microstructure, and creep properties, and offer a promising design strategy for developing economical high-temperature Fe-based alloys suitable for advanced applications.
36 MATERIALS SCIENCE↗
Development of Predictive Model for Accurate Rupture Time from Multi-Axial Creep in Alloy 709 with Physics-Based Simulations
A physics-based model is developed to predict multiaxial creep behavior in Alloy 709 (A709), an advanced austenitic stainless steel intended for high-temperature applications such as Sodium Fast Reactors (SFRs). Compared to conventional stainless steels like 316H, A709 offers superior high-temperature performance; however, comprehensive data on its multiaxial creep response remain limited. To address this gap, a crystal plasticity finite element (CPFE) framework is used to simulate the deformation and failure mechanisms of A709 under multiaxial loading conditions. The model incorporates an extended Hu-Cocks dislocation creep formulation that accounts for precipitation effects, along with the Sham–Needleman model to capture grain boundary cavitation-driven failure. These advanced constitutive models enable a detailed understanding of the interplay between microstructural evolution and macroscopic creep response. Furthermore, the study evaluates the predictive accuracy of various effective stress measures in estimating creep rupture life, leveraging simulated multiaxial creep data. The findings provide critical insights into the applicability of different stress measures for engineering design and life prediction of A709 components operating under complex loading conditions. This work contributes to improving the reliability of high-temperature structural components by advancing predictive modeling capabilities for advanced austenitic steels.
Alloy 709↗
Long-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at The LWR Relevant Temperature
Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer periods of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) initiated a collaboration with the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program on the Advanced Radiation Resistant Materials (ARRM) program, focused on developing and test degradation resistant alloys for light water reactor (LWR)- relevant environments. Based on a comprehensive microstructure and property screening, the ARRM program selected a total of five alloys (i.e., Grade 92, 310, 690, 718A, and 725), together with 316L and X-750 as references, for further investigations.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Status report on HFIR irradiation of optimized alumina forming alloys
Properties of FeCrAl alloys under neutron irradiation are of interest because of these materials’ potential application as accident-tolerant fuel cladding in nuclear systems. In parallel, alumina-forming austenitic (AFA) alloys are of interest for use as structural materials in advanced nuclear systems for their potential higher resistance to embrittlement and high-temperature steam oxidation resistance. An irradiation campaign for fiscal year 2024 has been developed under the Advanced Fuels Campaign to perform irradiation testing of various FeCrAl and AFA alloys in Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR). The goals of this irradiation campaign are to (1) study the impact of minor alloying elements on the neutron-irradiated mechanical properties of FeCrAl alloys and (2) collect neutron-irradiated mechanical properties on AFA alloys for comparison with those of FeCrAl alloys. This campaign will include both tensile and fracture toughness specimens tested following HFIR irradiation at temperatures representative of normal operating conditions in light-water reactors. The pre-irradiation characterization to date, the irradiation plan for the FeCrAl and AFA specimens, and the subsequent post-irradiation experimental test plan are presented in this report, along with the status of HFIR builds and scheduled insertion dates.
36 MATERIALS SCIENCE↗
Additive manufacturing of soft magnetic high entropy alloys: A review
Additive manufacturing (AM) offers unique advantages in fabricating soft magnetic high-entropy alloys (HEAs), enabling precise control over material properties and the development of advanced components for applications such as magnetic cores, electric motors, and transformers. These HEAs exhibit superior magnetic performance, mechanical strength, and thermal stability, making them highly suitable for modern electronics applications. Here, this review explores the advancements in AM-processed soft magnetic HEAs, including ongoing research on process optimization, tailored microstructures, and enhanced magnetic properties. It emphasizes the importance of understanding the correlations between AM process parameters, resulting microstructures, and the soft magnetic properties of HEAs. By summarizing the state of the field, we provide insights into current progress and highlight future research trends, focusing on the potential for industrial adoption and advancements in this emerging area.
36 MATERIALS SCIENCE↗
In-Situ Monitoring Assisted Large-Scale Additive Manufacturing of Mild Steel and 316L Alloys for Nuclear Application
The Advanced Materials and Manufacturing Technologies (AMMT) Program is aimed at developing cross-cutting technologies in support of a broad range of nuclear reactor parts, and to maintain U.S. leadership in materials and manufacturing technologies for nuclear energy applications. The overarching vision of the AMMT program is to accelerate the development, qualification, demonstration and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy. One of its three goals is to target big challenges and game-changing technologies, to realize the mission and vision of AMMT program. Based on this context, this multi-year work package focuses on understanding the current state of large-scale additive manufacturing (AM) technology for the deposition of 316L stainless steel materials for final components and mild steel for use in nuclear manufacturing processes. The targeted AM modality is directed energy deposition (DED), capable of fabricating components on the size scale of meters including valves, pumps, impellers. etc. that are challenging or difficult to source, especially when developing new systems or replacing obsolete components. Accordingly, the current writeup aims at providing a baseline literature survey on structure-property relationships in mild steel and 316L alloys. Also, information on preliminary trials to date involving these two alloys show tremendous potential of printing parts having complex geometry and thin- walled structures, such as nuclear valve and Hot isostatic Press (HIP) can, using wire based (Wire Arc Additive Manufacturing and Hybrid Additive Manufacturing) as well as blown powder DED machines. All of this is aimed towards (i) demonstrating the ability to fabricate large components for pressure boundary applications relevant to the nuclear community and nuclear manufacturing technology, and (ii) understanding the effect of different manufacturing technology on AM can production and post HIPed material for nuclear applications. Another target of this writeup is to compile various in-situ monitoring tools that have been incorporated for different DED AM modalities, in order to understand process variability during the entire fabrication process. This can be correlated with processing-structure-property response surfaces and would add confidence around process quality verification and ultimately component certification for nuclear applications.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Comparison of fluence-rate determination techniques for the Advanced Test Reactor-Critical Facility
A technique using small, enriched uranium-aluminum alloy monitors to measure fission rates in fuel plate assembly reactors was developed at the Idaho National Laboratory Radiation Measurements Laboratory in support of Advanced Test Reactor Critical (ATR-C) Facility irradiations. Previous measurements were performed to determine the correlation between the beta-particle activity and the fission rate in the U-Al monitors following irradiation. The American Society for Testing and Materials (ASTM) has established a standard practice E704 for determining neutron fluence rate by radioactivation of depleted uranium by measuring the concentration of fission-products in irradiated uranium monitors. Rather than measuring the beta-particle decay from fission-products, the decay rate of fission-products is determined by gamma-ray spectroscopy. The feasibility of using a method adapted from the ASTM standard practice E704 was investigated for ATR-C irradiations and compared to the present method by analyzing a subset of 21 monitor wires using both methods. The resulting calculated fission rates agreed with an average standard deviation of 7.45%, within the average combined measurement error of 9.05%. Practical considerations such as expected fluence-rates, sample activity, and counting-time were compared between the two methods and the sources of error for each method were investigated. Quality assurance measurements utilizing the modified ASTM standard method are recommended for future test irradiation, but gross-beta-particle counting remains the most practical solution for ATR-C fluence-rate determination.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
High Performance Alloy Technology to Market Support
The goal of the Technology to Market Assessment for Advanced Alloys is to identify non-technical support needs of research teams that could accelerate a path toward commercialization. The successful adoption of new technology into the market has positive impacts on direct stakeholders, taxpayers, and is important to showcase the strength of the United States (U.S.) Department of Energy (DOE) National Energy Technology Laboratory (NETL) Crosscutting High Performance Materials program. This assessment leverages a series of proven frameworks developed by NETL support contractor innovation specialists to determine which projects within NETL’s Crosscutting Advanced Alloys portfolio are in the best position for commercialization. The projects were then assessed for specific non-technical support needs.
29 ENERGY PLANNING, POLICY, AND ECONOMY↗
High-throughput design of high-performance lightweight high-entropy alloys
Developing affordable and light high-temperature materials alternative to Ni-base superalloys has significantly increased the efforts in designing advanced ferritic superalloys. However, currently developed ferritic superalloys still exhibit low high-temperature strengths, which limits their usage. Here we use a CALPHAD-based high-throughput computational method to design light, strong, and low-cost high-entropy alloys for elevated-temperature applications. Through the high-throughput screening, precipitation-strengthened lightweight high-entropy alloys are discovered from thousands of initial compositions, which exhibit enhanced strengths compared to other counterparts at room and elevated temperatures. The experimental and theoretical understanding of both successful and failed cases in their strengthening mechanisms and order-disorder transitions further improves the accuracy of the thermodynamic database of the discovered alloy system. This study shows that integrating high-throughput screening, multiscale modeling, and experimental validation proves to be efficient and useful in accelerating the discovery of advanced precipitation-strengthened structural materials tuned by the high-entropy alloy concept.
36 MATERIALS SCIENCE↗
Characterization of Alloy 709 Commercial Heats
The creep-resistant austenitic stainless steel Alloy 709 (Fe-20Cr-25Ni (wt%) based steel) is being investigated as a candidate structural material for the next generation of advanced reactors. Unlike conventional solid solution strengthened austenitic stainless steels, Alloy 709 develops a variety of precipitates during aging at different temperatures and times. In this study, precipitate evolution during short-term aging at 775°C was characterized. The effect of aging time (e.g., 10 vs. 100 hours) and cooling after aging (e.g., water quench vs. air cool) on grain size and precipitates have been investigated using light optical microscopy (OM), electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), and energy-dispersive X-ray spectroscopy (EDS). MX (Ni,Tb)(C,N) and M23C6 (Cr,Mo)C precipitates were observed in all the statically-aged samples. The grain size distribution in the sample aged for 10 hours and water quenched (10Q) versus the one aged for 100 hours and air cooled (100A) were comparable. This indicates that grain structure was fairly stable when aging at 775°C for up to 100 hours. In contrast, the sample aged for 10 hours and air cooled (10A) showed the smallest and the most uniform grain size distribution. This phenomenon was observed in only one sample. More studies on duplicate samples are needed to confirm the results obtained. The cooling medium (e.g., water quench vs. air cool) was not observed to impact the size nor the distribution of the MX and M23C6 precipitates. Increasing the aging time resulted in the following: (1) a decrease in dislocations as well as the number density of small precipitates; (2) an increase in the ratio of free precipitates that were not pinning dislocations; (3) relatively higher N, Nb, and Ti concentration in MX precipitate; (4) some increase in length of M23C6 carbides which were located on the coherent boundaries; and (5) the introduction of Sienriched M6X precipitates. The tensile tests at room temperature showed similar properties for each of the static aging treatments, indicating that the differences observed in the microstructural analysis did not have a significant impact on the tensile properties.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of New Reactor Core Configuration for Power Uprate - Fuel Reload & Heat Processing Analyses, Core Design, System Safety Assessments, and Fuel Performance Analyses
With the passage of the Infrastructure Investment and Jobs Act in 2021 and the Inflation Reduction Act (IRA) in 2022, the United States stands at a critical juncture for the future of nuclear power. These landmark policies provide significant support for clean energy initiatives, positioning nuclear power as a key component of the nation’s strategy to reduce carbon emissions and achieve energy security. This growing emphasis on nuclear energy is driven by the need for reliable, low-carbon power sources as the country transitions away from fossil fuels. Federal policy, along with increasing state-level support, is encouraging investment in nuclear technology advancements to meet these demands. Building new nuclear power plants (NPPs), however, presents significant challenges due to high costs and long construction timelines. As a result, increasing the power output of existing NPPs through power uprates has emerged as a more feasible and cost-effective strategy. One key area of advancement is the development of accident-tolerant fuel (ATF), such as chromium-coated zirconium alloy cladding, which offers enhanced material performance, enabling power uprates in light water reactors (LWRs). Given the growing demand for nuclear energy fueled by federal policies and state initiatives, it is essential to evaluate the feasibility and benefits of significant power uprates in existing pressurized water reactors (PWRs) using advanced fuel technologies. The introduction of ATF concepts opens new opportunities for safely and economically achieving these power increases. Assessing whether these innovations can support substantial power uprates while maintaining operational safety is crucial to maximizing the potential of the nation’s existing nuclear infrastructure. This project aims to explore how power uprates can be achieved by boosting reactor thermal power output and optimizing reactor core design, while ensuring the safety and economic viability of NPPs. Specifically, it will focus on demonstrating the technical and economic feasibility of power uprates in a PWR using low 5-10% enrichment uranium (LEU+) high burnup (HBU) fuel combined with ATF concepts. In fiscal year 2024 (FY24), the research and development focus on building foundational models and conducting multi-physics performance and safety analyses to support the power uprate. The findings of the study would be shared through LWRS Seasonal Meetings, conferences and workshops with utility companies and researchers. These also serve as a basis for further study of fuel reloading optimization with ATF claddings.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Advanced Ultra-Supercritical Component Test (ComTest) Project for 760*C Steam Conditions
ComTest is a $27M Department of Energy-funded project. Phase I, which began in November 2015, served to identify the technology gaps, as well as the scope and cost of required testing. Phase II, which was awarded in December 2018, includes an advanced manufacturing effort to complete U.S. based supply chain development for full commercial scale (800-850 MWe) AUSC components made of nickel-based alloys, components operating at up to 760°C. Completion of Phase II scheduled for September 30, 2021. Strategic Objectives of U.S. AUSC Project include: Power Plant Efficiency Improvements –Develop cost-effective, reliable technologies to improve the efficiency of new and existing high-temperature advanced generation power plants. Close gaps to achieve readiness for commercial scale demonstration of Advanced Ultra-Supercritical (AUSC) technology. Fabricate full-scale versions of key nickel-based alloy components. Validate capabilities of US supply chain for cost certainty. Support cross-cutting high-temperature generation technologies. Obtain ASME Code approval for new materials, components and processes. Increase power plant steam temperatures for higher cycle efficiency (Note: Average efficiency of US coal-fired fleet = 33% HHV; A-USC plant efficiency over 47% HHV at 1,400°F (760°C) steam temperature).
fireside corrosion↗
Mechanically Robust Bismuth-Embedded Carbon Microspheres for Ultrafast Charging and Ultrastable Sodium-Ion Batteries
Advancements in the development of fast-charging and long-lasting microstructured alloying anodes with high volumetric capacities are essential for enhancing the operational efficiency of sodium-ion batteries (SIBs). These anodes, however, face challenges such as declined cyclability and rate capability, primarily due to mechanical degradation reduced by significant volumetric changes (over 252%) and slow kinetics of sodium-ion storage. Herein, we introduce a novel anode design featuring densely packed bismuth (Bi) embedded within highly conductive carbon microspheres to overcome the aforementioned challenges. Remarkably, the high loading Bi anode within carbon microspheres with a high tap density of 2.59 g cm -3 possesses significant mechanical strength exceeding 590 MPa and limits volume swelling of only 10.9% post-sodiation. This anode demonstrates a high volumetric capacity (908.3 mAh cm -3 ), ultrafast chargeability (200 A g -1 , full charge/discharge in just 5.5 s), and outstanding cyclability over 12,000 cycles and maintains exceptional cycling stability even at -30 °C. The full cell paired with a Na 3 V 2 (PO 4 ) 3 cathode retains over 80% capacity after 600 cycles at 36 C, demonstrating a remarkable rate capability of 126 C (full charge/discharge in 28.6 s). Our comprehensive experimental evaluations and chemo-mechanical simulations shed light on the mechanisms underpinning the anode's superior performance. In conclusion, this development marks a significant advancement in the design of durable and fast-charging anodes for high-performance SIBs.
25 ENERGY STORAGE↗
Evaluation of printability methodologies and feasibility to down select LBPF steel and other materials for nuclear applications
In this work, new research is underway to produce alloys beyond well-known alloys with modified compositions that are better suited to be produced with additive manufacturing. These alloys can only possess improved radiation tolerance but also improved high temperature strength, creep resistance while also using elements that reduce the amount of activation after irradiation. This work package supports the vision and goals of the Advanced Materials and Manufacturing Technology (AMMT) program relevant to accelerate the development and deployment of advanced manufacturing processes. This work focuses on evaluating the methodology for designing alloys best suited for AM processes based on “printability”, literature research to evaluate and down select steel alloy systems. In this work, microstructural characterization and testing activities have been performed on certain steel alloys to support the efforts from collaborating laboratories during the evaluation and printing of the test articles. The final recommendation will be performed by all four participating laboratories.
316L↗
The Elevated-Temperature Cyclic Properties of Alloy 316L Manufactured Using Powder Metallurgy Hot Isostatic Pressing
According to developers at the GAIN 2019 microreactors workshop, advanced manufacturing (AM) is needed to fabricate microreactor components. A methodology has not been established to qualify AM for elevated-temperature nuclear construction. A potential approach is to use the wrought design methodology by establishing the AM properties as equivalent to or better than wrought. This investigation aims to understand the relationship between powder metallurgy hot isostatic pressing (PM HIP) and material characteristics for Alloy 316L as they relate to elevated-temperature component design and construction. The elevated-temperature cyclic behavior of Alloy 316L manufactured by PM HIP differs from wrought. Of particular significance is the reduced number of cycles to failure for the PM HIP Alloy 316L compared to wrought. Additional work to identify the microstructural feature that control the elevated-temperature cyclic behavior is needed.
36 MATERIALS SCIENCE↗
Atomistic modeling of lanthanide diffusion in refractory body-centered cubic molybdenum
Lanthanide fission products can strongly interact with candidate cladding alloys, but their transport properties in refractory metals remain poorly understood. Here, in this work, we investigate the atomic-scale diffusion behavior of La, Ce, Pr, and Nd in body-centered cubic (bcc) molybdenum, a potential candidate for advanced nuclear cladding. Self-consistent mean-field transport modeling is performed to evaluate the fission product transport and vacancy mobility, informed by first-principles and nudged elastic band calculations of vacancy formation energies, migration barriers, and solute–vacancy binding characteristics. Compared with bcc Fe, lanthanide solutes in bcc Mo exhibit slower tracer diffusion due to higher vacancy formation and migration energies. Furthermore, the calculations reveal that the influence of fission products on migration barriers in bcc Mo are not as extensive in range compared to bcc Fe. Among the studied lanthanides, La exhibits the strongest vacancy binding while also being the fastest diffuser in Mo. These findings highlight how refractory bcc alloys can reduce fission product infiltration, offering valuable insight into the development of durable cladding systems for advanced reactors.
36 - MATERIALS SCIENCE↗