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DOE OSTI2025

Investigation of Corrosion Behavior of Carbide Phase Strengthened Nickel‐Based Alloy in Molten Fluoride Salt

The corrosion behavior of a high creep strength carbide-phase strengthened Ni-based alloy in molten FLiNaK (LiF-NaF-KF: 46.5-11.5-42 mol%) salt in the temperature range of 700°C–750°C has been investigated as a part of the development of structural alloys for fluoride salt-based molten salt reactors (MSRs). The alloy composition was designed based on Hastelloy-N, but with the goal of improving creep strength. Cr depletion depth, a measure of corrosion, was observed to be single micrometers after several hundred hours of corrosion testing. Sequential corrosion testing involving testing of pre-tested samples in fresh salt coupled with SEM-EDS, scanning transmission electron microscopy (STEM) examinations and thermodynamic and kinetic modeling, suggest that the corrosion rate at the alloy-salt interface is governed by diffusion of elements from the alloy bulk to the surface. The carbide phases in the corrosion-tested sample microstructure were identified to be largely M 6 C-type Mo-rich carbides and MC-type mixed carbides. Atom probe tomography (APT) showed some partitioning of Cr and Ti to the carbide phase and showed the carbide phases to be stable at the salt-facing surface.

Gordon, Ryan [Univ. of Wisconsin, Madison, WI (Uni↗

DOE OSTI2026

Charged Defects in UO 2 Bulk and Surface: A First-Principles Study

Uranium dioxide (UO 2 ) is the primary fuel used in nuclear reactors. Under the extreme heat and radiation inside a reactor, this material inevitably develops defects in its crystal structure. To investigate the nature and behavior of these defects, DFT+U calculations were employed to investigate charged point defects in both bulk UO 2 and its most stable surface, the (111) plane. The formation of defects and their impact on the electronic structure were systematically examined. The results reveal that these defects introduce localized electronic states, alter magnetic behavior, and modify the structural properties. In general, such defects act as deep traps capable of capturing and retaining charge carriers. The stability of these defects depends strongly on the chemical environment and the position of the Fermi level. Surface defect calculations reveal that oxygen vacancies form more readily at the surface than in the bulk over a wide range of electron chemical potential, with subsurface oxygen vacancies being more stable than those in the top layer. Overall, the findings demonstrate how charged defects influence magnetism, transport, and stability in UO 2 , providing insights that may guide improvements in the safety and efficiency of nuclear fuel.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Cost- and Risk-Based Seismic Design Optimization of Nuclear Power Plant Safety Systems

Seismic analysis, design, and qualification of systems, structures, and components (SSCs) is a significant contributor to the capital cost of a nuclear power plant. To reduce capital costs of advanced nuclear power plants and make commercial nuclear energy more competitive, innovations are needed in their structural design and construction, and not just in the reactor core and associated systems. Seismic isolation has been identified as an important cost-cutting technology that enables standardization of equipment across various sites. This paper develops and demonstrates a cost- and risk-based seismic design optimization of a representative safety system in a nuclear power plant with the dual goals of minimizing overnight capital cost and meeting safety goals. The design optimization can also include component seismic isolation, in which case, the optimized design includes a set of equipment that needs to be seismically isolated to minimize capital cost. The open-source codes MASTODON and Dakota are used for seismic probabilistic risk assessment and design optimization, respectively. A generic nuclear facility with a safety system comprising SSCs that are common to nuclear power plants is considered for the demonstration of the design optimization and is assumed to be located at the Idaho National Laboratory site. Generic costs and seismic design cost functions are assumed for the SSCs of the safety system. The sum of the costs of the SSCs is minimized in the optimization process, while the risk of failure of the safety system is provided as a constraint. Furthermore, results show that the optimization process reduces capital costs significantly while automatically prioritizing the safety of SSCs that contribute most to the risk of the safety system.

42 ENGINEERING↗

DOE OSTI2026

Triso Analysis Tool For Predictive Source Terms

Source term modeling for TRi-structural ISOtropic (TRISO) fuel has been performed for previous reactor designs, but few are available in the open literature. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that can be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. To meet this need, the TRISO Analysis Tool for Predictive Source terms (TRISO-ATOPS) was developed. This model calculates the release of the key safety-important fission products by diffusion through the kernel, silicon carbide and graphite based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system for removing fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. From this initial condition, the model calculates the fission product release for any transient temperature profile, and the fission product releases can then be used to assess radiological dose to the workers and the public using conventional radiological dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy Advanced Gas Reactor TRISO fuel development program. The example cases in this work demonstrate the flexibility of the model

Stoyer, Benjamin [Idaho National Laboratory (INL),↗

DOE OSTI2020

High-Level Considerations for Access and Access Controls by Design

Nuclear power plants must be, by design and construction, robust structures and difficult to penetrate. Ideally, for security, the reactors would be sited underground, beneath a massive solid block, too thick to be penetrated by tools or explosives with all communications and power transfer lines also underground and fortified. Limiting access with difficult-to-penetrate physical barriers is going to be key for determining response and staffing requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Advanced Multiscale Microscopy Characterization of High Burnup LWR UO 2 Before and After LOCA Testing

Previous microscopy work within the campaign has observed increased porosity coupled with subgrain formation in the mid-radial region of the fuel. However, examinations of high burnup fuel taken from multiple reactors with different operating conditions have shown that this structure can move radially inward depending on the fuel operating conditions. It is currently theorized that inter- and intragranular fission gas bubble nucleation is precursory to the grain subdivision observed in the mid-radial and central regions of the fuel. Characterization of samples after loss-of-coolant accident (LOCA) testing has revealed that these restructured regions with a high density of bubbles and subgrains pulverize during the transient test. It appears that the increase in fission gas bubbles coupled with grain subdivision renders the fuel mechanically weaker during a LOCA transient and thus susceptible to fuel fragmentation relocation and dispersal (FFRD). Work this past fiscal year has prioritized understanding this behavior by using advanced microscopy to investigate fission gas behavior in the as-irradiated and post-LOCA state. Additional work has been performed to verify the theory described above by analyzing multiple post-LOCA optical micrographs and comparing that to as-irradiated microstructural data. This document reports progress in the post-irradiation characterization of high burnup nuclear fuel with emphasis on the restructured fuel regions, particularly how microstructural features influence FFRD and fission gas release behavior in LOCA conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Draft Critical Minerals Evaluation Report: Advanced Materials and Manufacturing Technology

Future energy production by sustainable clean energy technologies will be influenced by the availability and projected supply risks of critical minerals. In this report we discuss this context from the perspective of a sustainable nuclear energy technology and the benefits from the deployment of advanced manufacturing. In the past, the demand of critical minerals in the energy sector was minor, but in recent years renewable energy systems do force a substantial increase in the requirements for critical minerals which further induces pressure on the supply of critical minerals such as chromium, cobalt, nickel, niobium, tantalum, titanium, tungsten, vanadium, and zirconium for nuclear energy technology. The demand of critical minerals for nuclear technology is, besides beryllium and hafnium, minor and less than 1 % of the world supply. This alleviates the negative impact as the nuclear industry may not be able to provide the most compelling priority for the suppliers and to other policymakers looking at the overall needs. In this draft report, detailed background information is provided on the mineralogy of the identified critical minerals, supply chain risks followed by the discussion of specific minerals applicable to material relevant to nuclear energy technology. Briefly, the impact of advanced manufacturing on savings of critical minerals and electric energy is discussed and its effect to nuclear energy technology evaluated. This report will be followed with a final report providing a strategic vision for executable actions to implement the savings through advanced manufacturing, which will be further explored in the upcoming months. The Gen-IV relevant critical minerals will also be evaluated in the context of the AMMT programs material priorities as identified in the material score cards, to determine priorities for future actions. Material savings will become even more substantial when commercial Gen-IV reactor technology becomes available because of their needs for structural materials with higher content in alloying elements for enhanced high-temperature properties and improved corrosion resistance. The requirement for critical minerals in nuclear technology (in kg/MWe) will therefore significantly increase by 2050, even though its contribution to the worldwide electric energy production capacity will remain at about 5 %. The share of nuclear on world’s electric energy production is, on grounds of high utilization and availability 10.2 % and is expected to increase to 12 % by 2050. Advanced manufacturing could aid the deployment of Gen-IV nuclear technology since critical materials savings of up to 30 % seems possible, promoting nuclear energy as a true sustainable clean energy technology together with hydroelectric power.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo

Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 10 20 and 1.5 × 10 21 fissions per cubic centimeter. Swelling measurement of the specimen sizes studied here required development and assessment of new methods for volume determination before and after irradiation. Laser profilometry and X-ray computation tomography (XCT) were used to provide preirradiation characterization of samples to determine the error and applicability of each to determine swelling following irradiation. These outcomes are contextualized through use of BISON simulations performed to assess the predicted expansion of U-Mo fuels subjected to the irradiation conditions of this work. Use of existing BISON fuel performance models predicted a maximum of 7% swelling under the irradiation conditions of this study. Pre-irradiation characterization revealed the as-cast U-Mo fuel samples were uniformly large-grained fully cubic U crystals with small U-C/N bearing precipitates and pores distributed throughout. Samples were found to contain a bulk porosity between .4 and 3% because of the casting process. Local porosity in areas far from large, interconnected pores was found by Slice-and-View to be under .2%. Nanometer-sized precipitates rich in C and N were identified in all samples, likely because of impurities during the fabrication process. Dendritic bands were also observed throughout the samples. These bands were characterized by variable Mo content that deviated from the overall Mo content by 2–3 wt %. No other microstructural features were correlated to these bands. Mechanical properties were found to be slightly strengthened compared to literature reports of bulk U-Mo fuels due to the nano-scale precipitates throughout the sample.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Low Temperature Water-Gas Shift: Enhancing Stability through Optimizing Rb Loading on Pt/ZrO 2

Recent studies have shown that appropriate levels of alkali promotion can significantly improve the rate of low-temperature water gas shift (LT-WGS) on a range of catalysts. At sufficient loadings, the alkali metal can weaken the formate C–H bond and promote formate dehydrogenation, which is the proposed rate determining step in the formate associative mechanism. In a continuation of these studies, the effect of Rb promotion on Pt/ZrO 2 is examined herein. Pt/ZrO 2 catalysts were prepared with several different Rb loadings and characterized using temperature programmed reduction mass spectrometry (TPR-MS), temperature programmed desorption (TPD), diffuse reflectance infrared Fourier transform spectroscopy (DRIFTS), an X-ray absorption near edge spectroscopy (XANES) difference procedure, extended X-ray absorption fine structure spectroscopy (EXAFS) fitting, TPR-EXAFS/XANES, and reactor testing. At loadings of 2.79% Rb or higher, a significant shift was seen in the formate ν(CH) band. The results showed that a Rb loading of 4.65%, significantly improves the rate of formate decomposition in the presence of steam via weakening the formate C–H bond. However, excessive rubidium loading led to the increase in stability of a second intermediate, carbonate and inhibited hydrogen transfer reactions on Pt through surface blocking and accelerated agglomeration during catalyst activation. Optimal catalytic performance was achieved with loadings in the range of 0.55–0.93% Rb, where the catalyst maintained high activity and exhibited higher stability in comparison with the unpromoted catalyst.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

Tellurium Corrosion of Type 304/304L Stainless Steel, Iron, Chromium, and Nickel in High-Temperature Liquid Sodium

Investigating tellurium (Te) corrosion on structural materials is crucial for sodium-cooled fast reactors (SFRs) due to radionuclide presence and knowledge gaps. In this study, Type 304/304L stainless steel (SS304), chromium (Cr), iron (Fe), and nickel (Ni) samples were immersed in low-oxygen environments with Te in liquid sodium at 773 K for 30 days. At 10 ppm oxygen, SS304 showed multiple oxide layers, including a compact NaCrO 2 interlayer and porous Na-Fe-Ni-O outer layers. Tellurium penetrated through the porous layers but was hindered by the NaCrO 2 interlayer. At 0.01 ppm oxygen, Cr had no oxide layer, while Fe and Ni had unstable ones. Tellurium-induced pitting was deeper in Fe and Ni compared to Cr. Oxygen levels and Cr composition are critical factors affecting stable oxide compound layer formation and mitigating Te-induced pitting.

36 MATERIALS SCIENCE↗

NASA NTRS2000

Growth and Characterization of III-V Semiconductors for Device Applications

The research goal was to achieve a fundamental understanding of the physical processes occurring at the surfaces and interfaces of epitaxially grown InGaAs/GaAs (100) heterostructures. This will facilitate the development of quantum well devices for infrared optical applications and provide quantitative descriptions of key phenomena which impact their performance. Devices impacted include high-speed laser diodes and modulators for fiber optic communications at 1.55 micron wavelengths and intersub-band lasers for longer infrared wavelengths. The phenomenon of interest studied was the migration of indium in InGaAs structures. This work centered on the molecular beam epitaxy reactor and characterization apparatus donated to CAU by AT&T Bell Laboratories. The material characterization tool employed was secondary ion mass spectrometry. The training of graduate and undergraduate students was an integral part of this program. The graduate students received a thorough exposure to state-of-the-art techniques and equipment for semiconductor materials analysis as part of the Master''s degree requirement in physics. The undergraduates were exposed to a minority scientist who has an excellent track record in this area. They also had the opportunity to explore surface physics as a career option. The results of the scientific work was published in a refereed journal and several talks were presented professional conferences and academic seminars.

Williams, Michael D.↗

NASA NTRS2002

The FRC Acceleration Space Thruster (FAST) Experiment

The objective of the FRC (Field Reversed Configuration) Acceleration Space Thruster (FAST) Experiment is to investigate the use of a repetitive FRC source as a thruster, specifically for an NEP (nuclear electric propulsion) system. The Field Reversed Configuration is a plasmoid with a closed poloidal field line structure, and has been extensively studied as a fusion reactor core. An FRC thruster works by repetitively producing FRCs and accelerating them to high velocity. An FRC thruster should be capable of I(sub sp)'s in the range of 5,000 - 25,000 seconds and efficiencies in the range of 60 - 80 %. In addition, they can have thrust densities as high as 10(exp 6) N/m2, and as they are inductively formed, they do not suffer from electrode erosion. The jet-power should be scalable from the low to the high power regime. The FAST experiment consists of a theta-pinch formation chamber, followed by an acceleration stage. Initially, we will produce and accelerate single FRCs. The initial focus of the experiment will be on the ionization, formation and acceleration of a single plasmoid, so as to determine the likely efficiency and I(sub sp). Subsequently, we will modify the device for repetitive burst-mode operation (5-10 shots). A variety of diagnostics are or will be available for this work, including a HeNe interferometer, high-speed cameras, and a Thomson-scattering system. The status of the experiment will be described.

Martin, Adam↗

NASA NTRS1983

Ultrasonic Monitor

MicroUltrasonics PLR-1000 is a refined microprocessor-controlled version (usable on bolts, plates, liquids and gases) of the P2L2 developed by Langley Research Center. New technique is for nondestructive measurement of residual stress in various types of structures, for example, nuclear pressure vessels, pipes in nuclear reactors, offshore platforms, bridges, railroad tracks and wheels, aircraft wings, and engines. The instrument produces sound tone pulses that travel through a test specimen. PLR-1000 precisely measures speeds as stress increases speed of sound, tone changes, making precise measurements possible.

Source record↗

DOE OSTI2020

Integration of Data Analytics with Plant System Health Program

This report summarizes the R&D activities of the Plant Health Management (PHM) project during fiscal year 2020 (FY20). This project focuses on the development of methods that integrate component health data and propagate such information at the system level to evaluate system sources of risk. This project development lives in cooperation with the Risk Informed Asset Management (RIAM) project. The RIAM project, in fact, uses the reliability models and data generated by the PHM project to optimize plant operations (e.g., maintenance/replacement schedule, optimal maintenance posture). This year’s activities for the PHM project focused mainly on the development of two classes of models. The first one includes a series of component reliability models which include aging, testing and maintenance. The second class includes a series of system reliability models which focus on the secondary side of exiting U.S. reactors. These models are based on fault tree logic structures such that they can be used by existing plant PRA software. We also started to focus on the management of health data and we tackled this issue in two directions. The first one focuses on the integration of monitor data with simulation models to assess component health. This approach is moving from a classical data based to a model+data based approach with the goal of improving higher component health information. The second direction focuses on linking equipment reliability data (e.g., maintenance/failure reports, component monitoring data) directly to system reliability models using a safety margin based language rather than a probability based language. The main advantage of a safety margin based language is that it can provide to system engineers more tangible information on system/component health and how it propagates to the system level.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2022

ARC-100 Conceptual Design Phase Level 1 PSA Early Draft

This report represents the initial structure and content created for the ARC-100 nuclear reactor initial conceptual design Probabilistic Safety Assessment. The objective of the study is to assess the safety of the ARC-100 nuclear reactor conceptual design using a risk-informed approach. This early draft report is the first transmittal of the Probabilistic Safety Assessment approach and will be expanded over the remainder of the projects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Nuclear fuel rods and heat pipes in a graphite moderator matrix for a micro-reactor, with the fuel rods having fuel pellets in a BeO sleeve

A reactor unit cell is disclosed including a graphite moderator structure, a heat pipe positioned in the graphite moderator structure, and a fuel assembly positioned in the graphite moderator structure. The fuel assembly comprises at least one fuel rod. Each fuel rod comprises a beryllium-oxide sleeve and nuclear fuel positioned in the beryllium-oxide sleeve.

Levinsky, Alex↗

DOE OSTI2026

Multiscale modeling of packed-bed microwave reactors and estimation of intrinsic materials' permittivity

Modeling of packed-bed microwave reactors relies on an accurate representation of particle size, shape, and distribution within the bed, as well as the particles' dielectric properties. The measured permittivity of microwave susceptors (powders or structured materials) depends on the geometric features of the particles and the porosity of the bed, as well as the specific form factor of a structured material. These are effective properties and cannot be used to analyze other reactor configurations unless the geometric effects are removed. Therefore, we introduce a methodology for extracting the intrinsic particle permittivity from experimentally measured effective permittivity by combining cavity-based measurements with multiscale simulations and machine learning. Further, we develop the first multiscale model of packed-bed microwave reactors that incorporate particle effects (geometric features, random packing, and particle contact). This approach bridges macroscopic observables with mesoscopic physics, enabling analysis of local hotspots, arcing, and contact effects that control reactor performance. Using polymer-based spherical activated carbon (PBSAC) and silicon carbide (SiC) as examples, we demonstrate that the inferred particle permittivity is consistent with independent experimental heating profiles we collect from microwave reactors without adjustable parameters. Finally, this methodology establishes a foundation for predictive, multiscale design of microwave packed-bed reactors that explicitly accounts for particle-scale effects, enabling the estimation of intrinsic permittivity for the first time.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2020

Development of the MPACT 69-group Library for Magnox Reactor Analysis using CASL VERA

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnoxreactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

Kim, Kang Seog↗