High Temperature Mechanical Behavior of Laser Powder Bed Fusion of A709 and G92 Steels for Nuclear Applications
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Mantri, Srinivas↗
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Investigation of large-scale AM + PM parts for Nuclear Application
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Nag, Soumya↗
Intern Poster Session 08/13: Autonomous Nuclear Robotics: Applications in nuclear waste inspection and hot cell experiments
The nuclear industry is experiencing renewed interest in autonomous robotics, yet most deployed systems remain teleoperated with limited autonomy. This work presents two contributions toward fully autonomous nuclear robotic systems: autonomous waste inspection at the Hanford Site and an autonomous hot cell laboratory framework. Inspections of Hanford's underground waste storage tanks are performed manually at significant cost and personnel exposure. We developed a reinforcement-learning (RL) training pipeline for a custom-built inspection arm. In parallel, we are designing an autonomous laboratory framework for post-irradiation examination in hot cells at the Specimen Preparation Laboratory (SPL) that integrates computer vision, task and motion planning, hardware execution, and operator-in-the-loop control. These systems demonstrate a path toward safer, more efficient nuclear operations by reducing human exposure while maintaining rigorous human oversight at critical decision points.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Multi-Scale Modeling of the Evolution of Structure and Properties in Materials for Nuclear Energy Applications [Slides]
Nuclear energy is an important component of an overall strategy to address climate change. Idaho National Laboratory (INL) is the U.S. Department of Energy’s primary facility for research and development in nuclear science and technology for energy generation, supporting the improvement and life extension of the existing reactor fleet and the development and licensing of new reactor designs. Computational modeling is an important component of these activities, particularly in the area of materials for nuclear applications, where experimental data can be very challenging and expensive to acquire, and where data is especially scarce for new reactor designs. INL has used multi-scale modeling – linking atomistic, mesoscale, and engineering scales – to improve the ability to predict the performance of materials for nuclear energy applications. In this talk, I will give an overview of the approach and tools used, and several examples of application, including performance of nuclear fuels, understanding radiation-driven formation of nanoscale void and gas bubble superlattices, and powder densification through electric field assisted sintering.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
An Overview of Emerging Nuclear Sensor Technologies: Challenges, Advancements and Applications
Nuclear sensors are essential for detecting and measuring nuclear radiation in various applications, including nuclear power plants, medical imaging, and environmental monitoring. Traditional nuclear sensors have served these fields for decades, but recent advancements in emerging sensor technologies offer novel improvements in accuracy, sensitivity, and reliability. This review presents an up-to-date overview of recent progress in the advancements of nuclear sensor technologies, their diverse applications, challenges in implementation, and opportunities for future research.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Evaluation of additive manufacturing side channels for nuclear nonproliferation applications
The nuclear industry is continuing to grow and adopt new technologies to manufacture components. One of the technologies under consideration is additive manufacturing (AM). AM could make production of components at lower cost and could lessen lead times significantly. However, AM could be prone to proliferation risks. AM machines give off signatures during the manufacturing process, and these signatures offer an alternative means of observing the component. Correlations between these signatures and the geometry being manufactured could be developed; enabling proliferators as they could take advantage of these correlations to steal AM manufacturing instructions relating to nuclear technology. This work pertains to preliminary work exploring the feasibility of tapping into AM side channels to predict geometries being manufactured on AM machines. The feasibility of utilizing vibration to draw geometric correlations will be assessed on an Ultimaker 2 + thermoplastic 3D printer.
Additive manufacturing↗
A Review of the Research and Development of Brayton Cycle Technology in Nuclear Power Applications with a Focus on Compressor Technology
This study reviews the integration of Brayton Cycle (BC) systems in nuclear power generation, emphasizing their potential to enhance thermal efficiency and operational flexibility over traditional Rankine Cycle (RC) systems. Key working fluids, such as helium (He), supercritical carbon dioxide (sCO 2 ), nitrogen (N 2 ), and air, are evaluated for their performance, efficiency, and compatibility with nuclear systems. He is recognized for its high thermal conductivity and inertness at elevated temperatures, while sCO 2 demonstrates advantages in compactness and efficiency in midrange temperatures. This article also highlights the importance of compressor designs in optimizing BC performance and reviews, available compressor technologies. Axial and centrifugal compressor designs enable efficient gas compression while managing the thermal and mechanical stresses associated with high-pressure operations in nuclear systems. Combined with variable geometry components and advanced materials, these technologies address the challenges posed by varying load conditions. Despite the promising features of BC systems, several challenges persist, including high leakage rates and material degradation under extreme conditions, which necessitate robust sealing technologies and thorough testing. The insights gained from operational experiences at facilities, such as the Oberhausen II plant and the High-Temperature He Test Facility (HHV), underscore the complexities involved in designing high-temperature gas turbines for nuclear applications. This review concludes that as the nuclear industry evolves, BC systems hold significant promise for contributing to a sustainable energy future, particularly in the context of small modular reactors (SMRs) and microreactors. Further exploration of combined cycle configurations that combine BCs with RCs may enhance overall efficiency and flexibility in power generation.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Uncertainty estimation of bifurcated solutions in the Rayleigh–Bénard problem for advanced nuclear reactors applications
Multiphysics models of nuclear reactors frequently comprise nonlinear systems of equations. The nonlinear nature of these models could lead to solution bifurcations, where a small change in a certain parameter, e.g., the thermophysical properties of the coolant, can lead to a sudden change in the system’s behavior. At the point in parameter space where this happens, called a critical point, the Jacobian matrix of the model’s nonlinear operator becomes singular potentially permitting multiple solutions to coexist. In this paper, we perform uncertainty estimation (UE) in a parameter range that includes bifurcated solutions within the context of Rayleigh–Bénard problem. We perform this analysis assuming uncertain temperature difference, and tilt angle for the iterative solution algorithm with a unit Prandtl number (Pr = 1). Also, we perform this analysis under uncertain thermophysical properties for both FLiBe molten salt and liquid sodium as working fluid. We deploy two approaches to compute statistical moments for the resulting distributions of selected flow-field variables. The first approach is the blind computation of the mean and the standard deviation without any consideration of solution bifurcation, while the second approach utilizes k-means clustering to cluster each branch’s solutions together and compute separate statistical moments for each branch. The statistical distributions are obtained by perturbing the selected parameters about nominal values that correspond to a solution on one of the valid branches, and that solution is used as initial guess for the iterative solution algorithm. We found that perturbation of any parameter when its nominal value is close to its critical point always leads to branch jumping, i.e., the iterations converge to a solution on a branch different from the branch of the initial guess. This produces a statistical ensemble comprised of fundamentally different solutions leading to wrong mean values and uncertainty estimates, whereas clustering provides an efficient way to deal with this type of computation. This work is important for developing Gen IV nuclear systems because many of these systems rely on natural convection for cooling especially in accident conditions.
97 - MATHEMATICS AND COMPUTING↗
Bubbler Design Updates for Nuclear Safeguards Applications
The accurate monitoring of molten salt is crucial for nuclear safeguards, particularly in the context of mixtures used in molten salt reactors and pyropocessing. To attain material accountancy in actinide bearing molten salts, fluid volume and density are needed. Previous work utilized a triple bubbler sensor to measure the fluid level and density in processing tanks. This triple bubbler system utilized three gas dip-tubes strategically positioned at different heights within the molten salt such that the molten salt level, density, and surface tension were determined simultaneously. This investigation explores the effectiveness of an updated bubbler design with larger internal diameters for the dip-tubes to reduce plugging. The findings of this study will provide valuable insights into a bubbler design and configuration for applications in nuclear safeguards. This will contribute to enhanced accountability and transparency in the monitoring of actinide bearing molten salts, promoting objectives in nuclear safeguards.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Bubbler Design Updates for Nuclear Safeguards Application
The accurate monitoring of molten salt is crucial for nuclear safeguards, particularly in the context of mixtures used in molten salt reactors and pyropocessing. To attain material accountancy in actinide bearing molten salts, fluid volume and density are needed. Previous work utilized a triple bubbler sensor to measure the fluid level and density in processing tanks. This triple bubbler system utilized three gas dip-tubes strategically positioned at different heights within the molten salt such that the molten salt level, density, and surface tension were determined simultaneously. This investigation explores the effectiveness of an updated bubbler design with larger internal diameters for the dip-tubes to reduce plugging. The findings of this study will provide valuable insights into a bubbler design and configuration for applications in nuclear safeguards. This will contribute to enhanced accountability and transparency in the monitoring of actinide bearing molten salts, promoting objectives in nuclear safeguards.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Bubbler Design Updates for Nuclear Safeguards Applications
The accurate monitoring of molten salt compositions is crucial for nuclear safeguards, particularly in the context of mixtures used in molten salt reactors as well as pyroprocessing. However, for nuclear material accountancy in actinide bearing molten salts, the solution volume is also needed to determine the total mass of material in the salt. This study explores the effectiveness of two bubbler configurations, triple and single, for density and level measurements within these molten salts for salt volume determinations. The triple bubbler system utilizes three gas dip-tubes strategically positioned at different heights within the molten salt. With this design, the molten salt level, density, and surface tension can be determined simultaneously. The single bubbler configuration, on the other hand, relies on a dynamic single dip-tube to determine the density and level. Both configurations will be tested in aqueous solutions as well as in LiCl-KCl and NaCl-MgCl2 molten salts between 450°C and 750°C. Dip-tube diameters and bubbler inferences will also be explored. A new bubbler is being designed based on results. The findings of this study will provide valuable insights into the optimal bubbler design and configuration for ensuring representative sampling of molten salt compositions in nuclear safeguards applications. This will contribute to enhanced accountability and transparency in the monitoring of actinide bearing molten salts, promoting nuclear safeguards objectives.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
5G Communications in Nuclear: Potential Use Cases and Security Considerations
As fifth-generation (5G) communications continues to revolutionize the future of wireless technology, there is growing demand to utilize its benefits for critical infrastructures such as nuclear power plants (NPPs). In regard to achieving full automation and control in the operation of existing and future nuclear reactors, the unique capabilities of 5G can bring several potential advantages over other wireless technologies. However, a deep investigation is needed for the availability and security of 5G communications under various NPP operational scenarios. This article examines how 5G security capabilities can be architecturally deployed in nuclear applications so as to replace existing communication infrastructures. We discuss the current use of all wireless technologies in NPPs with their key features. Consequently, we investigated several NPP use cases in which 5G offers potential advantages but entails specific security considerations. The present article covers the characteristics of 5G communications, general challenges to its application in nuclear, and the security gaps that need to be addressed. We also highlight certain 5G security-by-design features that can help addressing current stringent NPP requirements. In addition, we discuss some future research direction that can facilitate the implementation of 5G in a nuclear facility. The findings presented herein can help foster 5G deployment in NPPs, thus enabling secured data transmission, cost savings, and increased operational efficiency with enhanced reliability.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Overcoming Challenges to Nuclear-Maritime Applications
This report is the fifth deliverable in a series of reports set forth by the Department of Energy (DOE), led by the American Bureau of Shipping (ABS), for the research award titled “Accelerating Commercial Maritime Demonstration Projects for Advanced Nuclear Reactor Technologies.” The report highlights guidance to bridge the gaps between innovative nuclear technologies and their practical implementation in maritime environments. The previous reports discussed specific technical, economic and regulatory challenges that may be expected when developing and establishing advanced nuclear technologies for commercial maritime applications. This report supports the mission of the U.S. National Reactor Innovation Center (NRIC) to demonstrate projects with industry and acts as a resource for the various stakeholders associated with the development of novel applications for advanced reactors. The previous reports discuss the interest in nuclear technology applications for decarbonized energy and identify the potential demands in nuclear energy supply and price. The work in these reports supports the recently issued nuclear Executive Orders (EOs). Specifically, EO 14299, “Deploying Advanced Nuclear Reactor Technologies for National Security”, and EO 14300, “Ordering the Reform of the Nuclear Regulatory Commission,” by ensuring the rapid development, deployment, and use of advanced nuclear technologies; and increasing the deployment of new nuclear reactor technologies, such as Generation III+ and IV reactors, modular reactors, and microreactors to support America leading the commercialization of affordable and abundant nuclear energy.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
𝛾-ray spectroscopy of 140 La decay for nuclear forensics applications
140 La is an important isotope in nuclear forensics. It is typically paired with 140 Ba as a chronometer for dating nuclear events. Here, in this work, we report the results of two measurements performed to fully characterize the decay scheme of 140 La. The experiments were performed using the Gammasphere array at Argonne National Laboratory and the CLARION2 array at Florida State University. Sources of 140 La were produced through neutron irradiation of natural lanthanum (99.9% 139 La) in the University of Massachusetts Lowell research reactor. We identify a total of 40 transitions, three of which were not previously reported. An angular correlation analysis was conducted confirming current literature 𝐽 assignments and mixing ratios. By considering all data available in the literature, we provide new recommended values for the intensities of the strongest transitions commonly used in applications.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Report on Potential Challenges and Impacts of Advanced Nuclear-Maritime Applications in the U.S.
This report is the third deliverable of a series of reports set forth by the Department of Energy (DOE), led by the American Bureau of Shipping (ABS), for the research award, “Accelerating Commercial Maritime Demonstration Projects for Advanced Nuclear Reactor Technologies.” The report highlights the challenges that are facing both the nuclear and maritime industry as technology seeks to integrate.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Ensuring Safe, Effective, and Reliable Use of Artificial Intelligence-Based Applications for Nuclear Electricity Generation through a Systems Approach
There is a significant potential to reduce operating and maintenance cost at nuclear power plants using artificial intelligence (AI) and machine learning (ML). For instance, AI/ML has the potential to significant improve work management processes, condition reporting, and plant surveillance activities. However, the nuclear industry has been slow in adopting AI/ML due to several multifaceted barriers discussed in this paper. This work presents such multidisciplinary approach intended to 1) accelerate industry adoption of AI/ML-based applications at nuclear power plants and 2) ensure their safe, reliable, and effective use. This approach is discussed further in this work and will be used to address overarching challenges associated with AI deployment at scale to provide industry guidance that support accelerated adoption of AI/ML technologies throughout the industry.
99 - GENERAL AND MISCELLANEOUS↗
The ENDF/B Nuclear Data Library and Its Impact on Reactor Simulations
The ENDF/B library, which is developed, maintained, and distributed by the Cross Section Evaluation Working Group, is the main source of nuclear data for analyses and computational simulations in nuclear applications, like different nuclear reactors concepts, radiation shielding, medical applications, astrophysics, etc. The library is constantly being improved and updated, with each release bringing an optimal representation of nuclear interactions as they are understood in their time. The most recent release, ENDF/B-VIII.1, represents a significant improvement in terms of the performance and consistency of the measured differential data relative to previous versions, as it combines the most recent experimental differential data and advanced theoretical nuclear models. As one of the many highlights, ENDF/B-VIII.1 restores a high-burnup depletion performance, comparable to ENDF/B-VII.1, that had been degraded in ENDF/B-VIII.0, while further improving the performance in criticality benchmarks, such as those in the Mosteller’s suite. Additionally noteworthy is the improved performance of ENDF/B-VIII.1 in radiation shielding and thick-target leakage spectrum integral experiments, which are also important for fusion and reactor applications. In this work we present in a very summarized way the main updates implemented in the ENDF/B-VIII.1 release and its main impacts, and also begin to delineate the path forward as to what to expect in the future for the next ENDF/B release, which, based on the timeline of the past few releases, is estimated to happen around 5 years from now. We emphasize that the ENDF/B-VIII.1 release was the product of an enormous collaborative effort among many authors and that for a complete detailed picture, the reader is strongly encouraged to refer to the article accompanying the release, which is currently in the publication process, but available as preprint [G. P. A. Nobre et al. arXiv Preprint arXiv: 2511.03564 (2025)].
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Influence of Powder Characteristics and Processing Methods on Creep Performance of Powder Metallurgy Hot Isostatic Pressed SS316
The U.S. nuclear energy expansion goals are driving the demand for manufacturing routes that can rapidly produce large, complex, near net shape components. Powder metallurgy hot isostatic pressing (PM HIP) is an advanced manufacturing technique that can be economically scaled-up, while alleviating the supply chain challenges that forging and casting face in terms of cost and lead time constraints. This makes PM-HIP a viable technology to aid and accelerate large-scale part manufacturing for nuclear applications. However, large-scale qualification and deployment of this technology require a thorough understanding of the influence of powder feedstock quality, powder handling history, hot isostatic pressing (HIP) parameters, and subsequent heat treatment on microstructural evolution and elevated temperature mechanical performance. The present work focusses on 316 austenitic stainless steel (SS316) which is most commonly used in high temperature environments for nuclear applications Results from this study show that PM HIPed SS316 meets ASME tensile requirements at room temperature and at elevated temperature. However, creep performance of PM-HIPed SS316 remains inferior to its wrought counterpart, demonstrating that tensile performance alone is not a reliable metric for long duration high temperature integrity. Further, this report delineates powder derived microstructural features that govern creep damage, with key evidences pointing to “microstructural inheritance” from gas atomized powder feedstocks. Commercial SS316 powders of varying chemical compositions and recycling histories were studies, and the results showed large differences in elemental segregation, oxide surface layers and secondary phase distributions. Multi-scale characterization revealed segregation of chromium, molybdenum, manganese and silicon at the boundaries and the precipitation of manganese-, silicon-, and molybdenum-oxides. During HIP consolidation, these surface oxides transform into decorated prior particle boundaries (PPBs) and grain boundary inclusions that persist through conventional post-HIP solution annealing treatment. The retained oxides in post-HIP microstructures were found to influence grain growth, precipitation behavior, and ultimately creep cavitation and fracture. Such post-HIP heat treatments are therefore limited by a complex trade-off between grain growth, and oxide coarsening which aggravate creep damage by acting as nucleation sites for cavities. The objective of this work is to establish an integrated processing–structure–property framework for PM-HIP 316 stainless steel by investigating the influence of powder feedstock characteristics in pre- and post-HIP processing as well as to understand the significance of post-HIP heat treatment on microstructural evolution and creep properties. The results from this report emphasize the significance of powder feedstock integrity in improving creep performance of PM-HIPed SS316, by highlight the effect of rapid solidification, elemental segregation, oxide formation and powder recycling on microstructural defect inheritance following HIP consolidation. Rather than considering HIP processing, solution annealing, and mechanical performance independently, this report treats powder production, HIP consolidation, post-HIP thermal processing, and creep deformation as interconnected stages within a continuous metallurgical process. The resulting framework will provide a scientific basis for developing feedstock engineering strategies capable of improving long-term reliability of PM-HIP stainless steels and accelerating their qualification for advanced nuclear applications.
Ajjarapu, Pavan [Oak Ridge National Laboratory (OR↗