Advanced Materials Program Summary
DOE Office of Nuclear Energy: Strategic Vision
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Summary report on the mechanical testing performed at INL for creep rupture, fatigue, creep-fatigue, and aging of A709 base material
The report provides the status of the creep, fatigue, and creep-fatigue testing to conducted at Idaho National Laboratory to generate a data package to support Alloy 709 qualification in American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 5. These tests were performed on three commercial heats of plate material. The first procured heat was manufactured by G.O. Carlson with heat number 58776. Second and third heats were fabricated by ATI Specialty Rolled Products, with heat numbers 529900 and 530843, respectively. A series of creep and cyclic specimens were fabricated from these three heats, and tests were performed. This report presents a master list of all completed and ongoing creep, fatigue, and creep fatigue tests. Cyclic properties of three commercial heats are compared. Thermal aging of three commercial heats for 3000 hours were completed in FY 2025. The thermal aging campaign to age plate material from three commercial heats up to 100,000 hours is ongoing.
36 - MATERIALS SCIENCE↗
Experiment design for the neutron irradiation of $\mathrm{PM-HIP}$ alloys for nuclear reactors
Here, this article describes the design of an Advanced Test Reactor (ATR) drop-in neutron irradiation experiment aiming to directly compare the performance of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP) against conventional casting or forging. There is growing interest in PM-HIP alloys for nuclear applications because of their microstructural uniformity, superior mechanical properties, and reduced dependence on welding and machining, compared to cast/forged alloys. Nuclear code-qualification of PM-HIP alloys requires neutron irradiation testing to demonstrate performance under relevant conditions. In this experiment, six nuclear structural alloys were irradiated: Ni-based alloys 625 and 690, Grade 91 ferritic steel, SA508 pressure vessel steel, and 304L and 316L austenitic stainless steels. The experiment is assembled into seven capsules in four test trains and irradiated in three ATR inboard A positions. Both the PM-HIP and cast/forged versions of each alloy were irradiated under nearly identical conditions for comparative purposes, to target doses of 1 ± 0.2 and 3 ± 0.2 dpa at temperatures of 300 ± 50 °C and 400 ± 50 °C. A thorough description of the experiment design and thermal, structural and neutronic analyses performed to ensure the targeted irradiation conditions are met is provided. Specimens were configured as small disks, compact tension specimens and tensile bars to facilitate post-irradiation examination (PIE) that will include mechanical testing, microstructure characterization, and fracture toughness testing. Given the considerations for ASTM standardized mechanical testing, comparative fluence and temperature across specimen pairs, and comprehensive PIE planning herein, this work serves as a template for future nuclear materials qualification experiment designs.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Demystify the 1.1 Factor for Ultimate Tensile Strength Above Room Temperature in Development of the Code Stress Tables
As the Boiler and Pressure Vessel Code increasingly gains popularity worldwide, it is not uncommon that many users are mystified by the 1.1 factor required for ultimate tensile strength above room temperature in developing the allowable design stresses of Section II Part D Stress Tables. Questions often arise about the origin and purpose of the factor, the reason for it not applied to the yield strength, its due considerations when evaluating an alloy for acceptance to construction use from the perspective of the alloy’s tested tensile strength value, and more. Not knowing reliable sources for explicit explanations, some users tend to misinterpret the factor, abuse its application, incorrectly define their alloy acceptance or qualification criteria, and mistakenly manipulate the safety margin in structural component design.To help the Code users who struggle with these frequent and confusing issues, particularly those in nuclear industry where rigorous criteria are required for component design and alloy acceptance or qualification, this paper is intended to demystify the 1.1 factor and facilitate knowledgeable interpretation and use of the Section II Part D Stress Tables as well as relevant Mandatory Appendices.A brief review is first given on the background of the 1.1 factor, followed by a summary of its application in the Stress Tables. The provenance and purpose of the factor are then discussed in detail with graphic examples. Finally, current applicability and necessity of the factor are considered through a demonstration using example alloys.
Ren, Weiju↗
Preliminary Experimental Results in Support of the Development of EPP and SMT Design Methods and Viscoplastic Model for A709
The ASME code qualification effort for Alloy 709 (A709) is currently underway to qualify it for Class A construction in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5. The United States (U.S.) Department of Energy (DOE) national laboratories are collaborating in the advanced materials development initiative to investigate the mechanical performance of A709 in support of its code qualification. As part of the A709 code qualification effort, this report summarizes ORNL’s initial experimental findings that support the integration of A709 into the elastic-perfectly plastic (EPP) strain limits code case. It also covers thermomechanical fatigue testing conducted to develop viscoelastic material models, along with the preliminary results of creep-fatigue experiments at 816°C using the Simplified Model Test (SMT) method.
36 MATERIALS SCIENCE↗
Draft ASME Boiler and Pressure Vessel Code Cases and Technical Bases for Use of Alloy 617 for Constructions of Nuclear Component Under Section III, Division 5
The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code currently only allows five materials for use in construction of nuclear components for high temperature service. These are: 2.25Cr-1Mo and V-modified 9Cr-1Mo steels, Types 304 and 316 stainless steels and the high nickel Alloy 800H. Since 2005, the US high temperature gas-cooled reactor program has been characterizing elevated temperature mechanical properties of Alloy 617 as the leading candidate construction material for the intermediate heat exchanger. After analysis of these experimental results, along with historical data and additional results available through the Generation IV International Forum, Very High Temperature Reactor, Materials Program Management Board Materials Handbook, a draft ASME Code Case to allow nuclear construction with Alloy 617 for temperatures up to 1750°F (954°C) has been developed. This report contains the Code Case for Low Temperature Service Construction of Section III, Division 5, Subsection HB, Subpart A, Class A and Subsection HC, Subpart A, Class B components, which has been approved in Section II, Materials, and Section III, Rules for Construction of Nuclear Facility Components. Supporting technical justification for the low temperature Code Case is also included. This Code Case allows use of Alloy 617 up to 800°F (425°C). This report also contains an updated draft of a Section III, Division 5, Subsection HB, Subpart B, Class A Code Case for Alloy 617 to qualify it for use in construction of nuclear components up to 1750°F (954°C) for service life up to 100,000 hours. The draft contained in Appendix 4, subject to editorial revision and approval by the ASME Special Task Group on Alloy 617 Code Qualification, will be submitted for approval by letter ballot by the appropriate ASME Committees. The technical justification supporting the Code Case is presented in Appendix 5 of this report. This background document is part of the information package that will be submitted with the Code Case for ballot.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Parameter, Post-Processing Sensitivities, and Qualification Approach of Laser Powder Bed Fusion Hydrogen Resistant Alloy NASA HR-1
Metal additive manufacturing (AM) processes are being used to enable economical manufacturing of legacy alloys as well as advancing new alloys. Laser powder bed fusion (L-PBF) is a metal AM process that has high maturity and being used to produce a variety of parts for space applications including complex propulsion components. The National Aeronautics and Space Administration (NASA) has identified the need to develop and advance new materials in unique space applications such as high-pressure hydrogen environments. NASA HR-1 is a high strength Fe-Ni based superalloy designed to resist high pressure hydrogen environment embrittlement (HEE), oxidation, and corrosion that has been successfully adapted to laser powder directed energy deposition (LP-DED). Insights gained from the NASA HR-1 development for LP-DED have guided the development process for L-PBF. However, adapting NASA HR-1 to L-PBF posed new challenges due to the distinct differences between the additive manufacturing processes. During parameter development, sensitivities were observed in post-processing that necessitated additional optimization of heat treatments. Additionally, the variations in thickness and how it influenced the microstructural response during heat treatment was characterized. Understanding these sensitivities is important to qualification of the material in a L-PBF machine. This ensures that the microstructures and properties of the material maintain consistency in production. This presentation will cover parameter development along with post-processing challenges and solutions will be discussed in addition to key material properties as it pertains to application performance and qualification per NASA-STD-6030. Improvements made by developing a derivative alloy, NASA HR-2, will be highlighted through preliminary small scale parameter development, material characterization, and initial property testing.
NASA HR-1↗
Interim Mechanical Properties Data from FY2022 ORNL Testing of A709 with Precipitation Treatment for ASME Code Case Data Package
A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being conducted at the US Department of Energy’s Oak Ridge National Laboratory, Idaho National Laboratory, and Argonne National Laboratory. A recent assessment on the Alloy 709 development and testing effort concluded that the mechanical properties of Alloy 709 with the precipitation treatment continued to outperform those of Type 316 stainless steel. The assessment also affirmed the recommendation for its Code qualification. This report provides the detailed integrated data generated to date on the two commercial heats of Alloy 709 in plate product form with precipitation treatment condition that were used as the basis for this assessment. Oak Ridge National Laboratory has continued to perform a subset of the Code Case testing for tensile properties, creep rupture, fatigue, and creep-fatigue. This report also updates the key Alloy 709 Code Case testing status and results in FY 2022 at Oak Ridge National Laboratory.
36 MATERIALS SCIENCE↗
Summary of FY 2024 A709 Code Case testing at ANL, INL and ORNL
A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being carried out at Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Key testing data for the Alloy 709 100,000-hr near-term Code Case submittal to ASME is expected to be completed by the end of 2024, with design parameters anticipated to be finalized in FY 2025. This report summarizes the testing results for three commercial heats of Alloy 709 conducted across three laboratories, reviews the current testing status, and outlines the remaining data needed to support the first Alloy 709 Code Case submittal to ASME. The Alloy 709 Code Case plan remains on schedule.
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Progress Report on Alloy 709 Base Metal Code Case Testing at ORNL in FY 2023
A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being carried out at Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL) and Argonne National Laboratory (Argonne). In FY 2023, ORNL has continued to conduct a subset of the Code Case testing for tensile, creep rupture, fatigue and creep-fatigue on Alloy 709. This report also updates the key Code Case testing status and results on the first two commercial heats and the preliminary results on the third commercial heat of Alloy 709. The three commercial heats of Alloy 709 are all in plate product form.
36 MATERIALS SCIENCE↗
Molten Salt Corrosion Tests of Additively Manufactured Stainless Steel 316H
Molten salt reactors (MSRs) have drawn considerable interest due to their favorable safety features, high thermal efficiency, and compatibility with different fuel cycles. Yet, the success of MSRs hinges critically on the performance of structural materials to be used in these aggressive molten salt environments, where corrosion and material compatibility remain primary challenges to long-term reliability. Additively manufactured (AM) nuclear structural materials prompt the use of novel geometries and compositions to enhance material performance and reduce costs of constructing MSRs. The rapid solidification conditions inherent to AM processing impart distinctive microstructural features, including cellular sub-structures, dislocation densities, residual stress, and oxide inclusions, which can influence material performance in MSR components. While the mechanical properties of AM stainless steels have been widely studied, their corrosion behavior, particularly in molten salt environments, has received far less attention. Addressing these needs, the Advanced Materials and Manufacturing Technologies (AMMT) program provides a framework for systematically evaluating how unique microstructures produced by AM processes influence the performance of these materials in these demanding environments and for developing reproducible testing workflows that can support future code qualification efforts and standards development. Bridging this knowledge gap is essential for assessing the viability of AM alloys in MSRs and informing qualification strategies. A further challenge is the absence of standardized protocols for molten salt corrosion testing. Accordingly, this report provides an account of the corrosion evaluation of AM 316H stainless steel in NaCl 2 -MgCl 2 molten salt at 550 °C, with exposure times of 100 and 500 hours. It documents the experimental procedures implemented under the AMMT program, including salt preparation, exposure protocols, and post-test characterization methods, to establish reproducibility and transparency. Importantly, the study examines AM 316H samples in the as-fabricated condition, directly reflecting the surface state most relevant to engineering applications, and compares their behavior to machine-cut surfaces. Overall, preliminary evaluations have noted that surface conditions (e.g. morphology, contamination, etc.) have a noticeable impact on the corrosion resiliency. The impact of the corrosion is difficult to detect at 100h, unless, in the case of AM 316H, the specimen surface is decontaminated. After 500 h, as-fabricated surfaces of AM and wrought 316H display evidence of general versus preferential corrosion attack, respectively. Both AM and wrought 316H machine-cut surfaces exhibit a continuous Cr depletion zone, evident of general corrosion. While the estimated extent of corrosion appears within the same order of magnitude regardless of the surface condition, it is apparent that more predictable behavior is observed on machine-cut surfaces. Nonetheless, further investigation is necessary to fully elucidate the corrosion mechanism under these conditions.
36 - MATERIALS SCIENCE↗
Laser Powder Bed Fusion Additive Manufacture Nb1Zr Development
Next generation fission and fusion nuclear reactors require materials that can withstand operating temperatures greater than 500 °C, neutron irradiation doses of up to 200 displacements per atom (dpa), and potentially corrosive coolants such as the alkali liquid metals sodium, lithium, and NaK (Na33K eutectic alloy). Refractory alloys, such as Nb1Zr (Nb-1wt%Zr) and Molybdenum alloy TZM (Mo-0.5wt%Ti-0.08wt%Zr) have been traditionally considered viable candidates for advanced fission and fusion reactor concepts. However, it is relatively difficult to generate complex geometries of interest from these alloys using traditional manufacturing methods. In addition, there needs to be a concentrated effort to address refractory metal challenges at elevated temperature operation. In order to generate complex geometries of interest, modern manufacturing techniques are considered to increase the technological readiness level (TRL), cost-effectiveness, and schedule savings. This work focused on the continued development of laser powder bed fusion (L-PBF) additive manufacturing (AM) to improve both design flexibility, evaluate microstructure and properties, and ultimately accelerate the TRL and qualification of these processes and alloys for components to potentially be put into service. Niobium alloy Nb1Zr was identified through a down-selection process outlined in previous reports as a candidate to develop in L-PBF AM. Historically, Nb1Zr had been explored for high temperature fast spectrum fission reactors for both terrestrial and space applications. Molybdenum alloy TZM has also been considered for these reactor concepts due to exceptional high-temperature strength, creep resistance, and stability under irradiation. L-PBF AM of TZM has previously been investigated at LANL under the Microreactor program, NASA, ORNL, and in academia. However, due to the crack prone nature of TZM, L-PBF AM of TZM resulted in significant microcracking and additional development is required to pursue viable maturation. Other AM methods have been found to be more successful in printing TZM, and those alternatives approaches are discussed in this effort. The efforts detailed in this report focused on continued development of Nb1Zr through L-PBF and development of TZM via L-PBF and electron powder bed fusion (E-PBF). The objective of this work was to further the development of these AM techniques for the chosen refractory alloys, elucidating and addressing associated challenges through characterization of several demonstration builds. At LANL, Nb1Zr builds were completed using an EOS M290 and M400 machines, and a refractory alloy-dedicated L-PBF system, the Xact Metal XM200G, was installed. The XM200G primary purpose was to do the Nb1Zr parameter development process; however, due to difficulties associated with the machine installation and qualification process, it was decided to pivot development to the larger M400 and M290 machines. Although the supply of Nb1Zr powder was limited, it was sufficient to generate sub-scale metallographic specimens for the purpose of parameter development. This was first accomplished on the EOS M400 then the M290 due to machine schedule availability. Further development of TZM has been initiated at the University of Texas El Paso (UTEP) under contract with LANL to use both a heated build envelope L-PBF machine and E-PBF machine that have been found in the literature to mitigate microcracking. UTEP was provided with TZM powder and build plates to support parallel TZM parameter development across both machines. As part of the contract, UTEP will also be conducting microstructural characterization once optimized process parameters have been identified. The optimized process parameters for each machine will be used to generate a series of metallographic, mechanical, and surface finish specimens for subsequent characterization and testing. In the next section, we provide a detailed discussion of the methodology used for investigating the feasibility of leveraging these alloys for use in advanced reactor applications.
36 MATERIALS SCIENCE↗
Comparing structure-property evolution for PM-HIP and forged alloy 625 irradiated with neutrons to 1 dpa
The nuclear power industry has growing interest in qualifying powder metallurgy with hot isostatic pressing (PM-HIP) to replace traditional alloy fabrication methods for reactor structural components. But there is little known about the response of PM-HIP alloys to reactor conditions. This study directly compares the response of PM-HIP to forged Ni-base Alloy 625 under neutron irradiation doses ~0.5–1 displacements per atom (dpa) at temperatures ranging ~321–385 °C. Post-irradiation examination involves microstructure characterization, ASTM E8 uniaxial tensile testing, and fractography. Up through 1 dpa, PM-HIP Alloy 625 appears more resistant to irradiation-induced cavity nucleation than its forged counterpart, and consequently experiences significantly less hardening. This observed difference in performance can be explained by the higher initial dislocation density of the forged material, which represents an interstitial-biased sink that leaves a vacancy supersaturation to nucleate cavities. These findings show promise for qualification of PM-HIP Alloy 625 for nuclear applications, although higher dose studies are needed to assess the steady-state irradiated microstructure.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Lessons Learned in the Flight Qualification of the S-NPP and NOAA-20 Solar Array Mechanisms
Deployable solar arrays are the energy source used on almost all Earth orbiting spacecraft and their release and deployment are mission-critical; fully testing them on the ground is a challenging endeavor. The 8 meter long deployable arrays flown on two sequential NASA weather satellites were each comprised of three rigid panels almost 2 meters wide. These large panels were deployed by hinges comprised of stacked constant force springs, eddy current dampers, and were restrained through launch by a set of four releasable hold-downs using shape memory alloy release devices. The ground qualification testing of such unwieldy deployable solar arrays, whose design was optimized for orbital operations, proved to be quite challenging and provides numerous lessons learned. A paperwork review and follow-up inspection after hardware storage determined that there were negative torque margins and missing lubricant, this paper will explain how these unexpected issues were overcome. The paper will also provide details on how the hinge subassemblies, the fully-assembled array, and mechanical ground support equipment were subsequently improved and qualified for a follow-on flight with considerably less difficulty. The solar arrays built by Ball Aerospace Corp. for the Suomi National Polar Partnership (S-NPP) satellite and the Joint Polar Satellite System (JPSS-1) satellite (now NOAA-20) were both successfully deployed on-obit and are performing well.
Helfrich, Daniel↗
Lessons Learned in the Flight Qualification of the S-NPP and NOAA-20 Solar Array Mechanisms
Deployable solar arrays are the energy source used on almost all Earth orbiting spacecraft and their release and deployment are mission-critical; fully testing them on the ground is a challenging endeavor. The 8 meter long deployable arrays flown on two sequential NASA weather satellites were each comprised of three rigid panels almost 2 meters wide. These large panels were deployed by hinges comprised of stacked constant force springs, eddy current dampers, and were restrained through launch by a set of four releasable hold-downs using shape memory alloy release devices. The ground qualification testing of such unwieldy deployable solar arrays, whose design was optimized for orbital operations, proved to be quite challenging and provides numerous lessons learned. A paperwork review and follow-up inspection after hardware storage determined that there were negative torque margins and missing lubricant, this paper will explain how these unexpected issues were overcome. The paper will also provide details on how the hinge subassemblies, the fully-assembled array, and mechanical ground support equipment were subsequently improved and qualified for a follow-on flight with considerably less difficulty. The solar arrays built by Ball Aerospace Corp. for the Suomi National Polar Partnership (SNPP) satellite and the Joint Polar Satellite System (JPSS-1) satellite (now NOAA-20) were both successfully deployed on-obit and are performing well.
Sexton, Adam↗
Focused Ion Beam Tomography of Alloy 617 Corroded in Molten Chloride Salt
Materials qualification of reactor structural materials is a critical step in rapid implementation of advanced nuclear reactor technologies, particularly to assess the corrosion performance in these designs. Accelerated qualification of reactor structural materials requires incorporating powerful computational toolsets, such as phase field modelling in the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, to predict the evolution of structural materials due to corrosion. Accordingly, computational toolsets will require experimental data generated at appropriate length scales to validate accuracy. Focused ion beam (FIB) provides a high degree of control over manipulation of materials for analytical purposes, including capturing data on the evolution in the microstructure and elemental composition of materials at the mesoscale, an appropriate length scale for phase field modelling of intergranular diffusion phenomena using the MOOSE framework. For instance, the FEI Helios G4 UX dual beam plasma FIB microscope at the Irradiated Materials Characterization Laboratory (IMCL) is capable of backscatter diffraction (EBSD) and energy-dispersive x-ray spectroscopy (EDS) documenting the evolution in the microstructure and elemental composition, respectively. The Helios can perform EDS and EBSD three-dimensionally (3D) using tomography, which is then combined using different software packages to visualize 3D volumes correlating elemental composition to microstructural data. The purpose of this investigation was to develop a streamlined characterization and data processing workflow for 3D tomography studies on the FEI Helios G4 plasma FIB. The investigation is segmented into three parts: 1) Optimizing the data collection workflow, 2) identifying appropriate data processing and visualization software (i.e. DREAM.3D, MIPAR, and VGStudioMax), and 3) establishing an infrastructure for public release. The optimization of the data collection workflow is in collaboration with members of the U220 department to setup formal training on the tomography operation of the G4, through ThermoFisher Scientific, and exploring DREAM.3D, MIPAR, and VGStudioMax data processing/visualization software packages. VGStudioMax currently demonstrates the most promise for future use. Optimization of the data collection and processing workflow is still ongoing. A collaboration with INL High Performance Computing (HPC) established an open-source license for expediting the public release of FIB tomography datasets through HPC. FIB tomography data generated by the G4 will provide comprehensive data for validating 3D phase field mesoscale modelling tools within the MOOSE framework for accelerated qualification of reactor structural materials.
Copeland-Johnson, Trishelle↗
High Temperature Testing of Refractory Alloys
- NASA Refractory Alloy Projects: - STMD-GRC – Refractory Alloy Additive Manufacturing Build Optimization (RAAMBO) - Hypersonic Technology (HTP) Project. - Advance material characterization and prototypic testing allowing for critical data needed to demonstrate TRL advancement and for component flight qualification - Enable national supply chain for refractory alloy feedstock essential for production of historical to custom alloys for R&D to component fabrication. - Develop materials and processing for high performance in extreme environments enabling technologies for Moon to Mars missions.
Justin L. Milner↗
Chrome - Free Aluminum Coating System
This slide presentation concerns the program to qualify a chrome free coating for aluminum. The program was required due to findings by OSHA and EPA, that hexavalent chromium, used to mitigate corrosion in aerospace aluminum alloys, poses hazards for personnel. This qualification consisted of over 4,000 tests. The tests revealed that a move away from Cr+6, required a system rather than individual components and that the maximum corrosion protection required pretreatment, primer and topcoat.
Bailey, John H.↗