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At least 703 records · Page 39
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

SPC-71123 Rev 1 MARVEL Control Drum Actuator Torsional Spring

Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core under the Microreactor Application Research Validation and Evaluation (MARVEL) program. This specification defines the requirements for the procurement and factory acceptance testing of the MARVEL control drum actuator torsional spring. The purpose of this specification is to define the requirements to procure the MARVEL control drum actuator torsional spring component and witness a torsional spring rate Factory Acceptance Test (FAT) at the subcontractor’s facility with the contractor present. Any conflict between this specification and referenced Codes and Standards, or any supplementary specifications in the procurement documents requires written clarification from the Contractor prior to proceeding with any work. Any deviation from the procurement documents requires approval by the Contractor with the change request process.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

Effect of iodides on thermal behavior and phase partitioning in LiCl-KCl

Liquid-fueled molten salt reactors (MSRs) are designed to operate with fissile materials and, ultimately, fission products dissolved in the primary molten salt coolant. Understanding the speciation and transport of iodine—a high-yield fission product—is essential because this element’s accidental release poses significant environmental concerns due to its capacity to be readily absorbed by the human thyroid gland. Here, we report the impact of iodide species (LiI and KI) on phase transitions, phase distribution, and phase stability in LiCl–KCl-eutectic salt mixtures. The study employed a combination of computational and experimental techniques, including thermodynamic FactSage calculations, differential scanning calorimetry, and high-temperature X-ray diffraction. The results indicate that the presence of iodide (10–25 wt%) significantly alters the melting behavior of the LiCl–KCleutectic system. Adding 10 wt% LiI has a more-pronounced effect than 10 wt% KI, as LiI converts to KI, leading to formation of LiCl, thereby, altering the LiCl-KCl ratio which significantly affects the melting temperature of the mixture. Furthermore, the evolution of crystalline structure, solid-fraction composition, and the dynamics of mixed-halide solid–liquid partitioning as a function of temperature indicate the potential for selective iodide separation from chloride-salt mixtures via solid–liquid separation techniques. Overall, the presented findings provide valuable insights that are beneficial for the design and operation of MSRs, as well as for the safe handling and effective processing of used nuclear fuel using advanced pyrochemical techniques.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Preliminary Evaluation of Loading DOE Standard Canisters in the INL CPP-603 Irradiated Fuel Storage Facility - 20543

This paper looks at the equipment and operations necessary to load United States Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) into DOE Standard Canisters in the CPP-603 Irradiated Fuel Storage Facility (IFSF) in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The fuel ready for loading would come from fuel storage canisters in the CPP-603 facility. The paper describes the facility, the fuel types, the DOE Standard Canisters, and existing equipment; lists the needed loading operations; reviews facility features and equipment to perform the operations; and then lists the decisions, analyses, designs, demonstrations, and modifications that will be needed to perform the loading of DOE-owned SNF into DOE Standard Canisters in the CPP-603 IFSF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

TRISO Fuel’s Safety Functions, Contributions to Reactor Safety, and Necessary Safety Limits

Safety functions are the actions, passive or active, that structures, systems, and components of nuclear facility that contribute to the safety of the workers, the public, or the environment. Well-defined safety functions are the foundation of a solid safety case for a reactor. For reactors that use TRISO-coated particles, the TRISO fuel plays an important part of the safety case because of its ability to contain radionuclides in the fuel itself. This ability enables the use of a functional containment strategy for the reactor where radionuclide retention is the primary safety function supported by the safety functions of controlling reactivity control and controlling heat rejection. This paper establishes at a deeper level the role that TRISO fuel plays in each of these safety functions and associated quality assurance and testing requirements for TRISO particle manufacturing to ensure these safety functions. Safety limits necessary to protect these safety functions include: operational limits, time at temperature limits, and fission gas release activity limits. In conclusion, this approach demonstrates the role that specific aspects of TRISO fuel play in protecting the safety of workers, the public, and the environment.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Non-destructive electrochemical diagnosis of failure mechanisms in aqueous zinc batteries

The early detection of secondary reactions that affect the life and performance of zinc manganese oxide batteries requires a shift from conventional time-consuming and often destructive procedures to rapid lifetime-predictive techniques. In this work, an electrochemical approach is employed to elucidate independent signatures for four common types of failure mechanisms in zinc manganese dioxide (Zn||MnO2) batteries—namely, the loss of zinc inventory, the loss of active material at the cathode, electrolyte depletion, and increased cell impedance. Our findings, specific to coin cell configurations, reveal that each induced failure mechanism can be distinctively modeled and identified based on responses from the rest voltage and columbic-efficiency data for prompt detection. For instance, electrolyte depletion response manifests a distinctive abrupt (>80 %) decrease in columbic efficiency (CE) and charge-rest voltage (Vc) while the discharge-rest voltage remained constant at ~1.3 V. Furthermore, electrolyte rejuvenation of the cell increased the CE to >95 % and restored Vc from ~0.3 to >1.7 V. Recovery experiments and reference performance tests demonstrated consistency between electrochemical descriptors and their associated failure mechanisms. Further, the outcomes of this work provide valuable insights and data models for some of the dominant failure mechanisms present in zinc manganese battery chemistries, which are beneficial to accelerated early-lifetime diagnosis and advancement of Zn batteries development.

25 ENERGY STORAGE↗
DOE OSTI2025

Technical Assessment of the Application of Digital Twin and Prognostic Tools for Condition Monitoring

This report was prepared for the U.S. Nuclear Regulatory Commission (NRC) to present use cases of the application of advanced technologies toward meeting the current and future regulatory requirements for maintenance and condition monitoring of structures, systems, and components (SSCs). The advanced technologies considered in this work, collectively referred to as digital twin (DT) technologies, are advanced sensors and instrumentation, data analytics, machine learning and artificial intelligence (ML/AI), and physics-based models. The report presents two use cases of reactor coolant pumps (RCPs) and heat pipes in nuclear power plants (NPPs) with technical and regulatory considerations and opportunities in using advanced technologies for conditional monitoring. Key findings from the exploration of these considerations are as follows: - Uncertainties in sensor data and model predictions must be rigorously addressed through validation and verification processes - Regulatory compliance is paramount, necessitating data driven models to be developed in line with existing codes and standards, as well as considering potential future guidelines for advanced reactors - Explainability and transparency in ML/AI models are essential for developing operator trust and regulatory review, including methods that enhance the interpretability of complex data-driven predictions - Condition monitoring programs must be evaluated for their effectiveness in reducing maintenance-preventable function failures (MPFF) and aligning with plant performance criteria - The deployment of advanced technologies for condition monitoring could lead to a transition from periodic to continuous monitoring, thereby optimizing maintenance schedules - Collaborative efforts between industry stakeholders, regulatory bodies, and technology developers are crucial for the successful adoption of advanced technologies for condition monitoring systems in nuclear facilities In summary, the introduction of advanced technologies into condition monitoring programs represents a significant leap forward in the domain of NPP maintenance. By harnessing the capabilities of advanced sensors, data analytics, and ML/AI, NPP operators can transition from a time-based to a condition-based maintenance approach. This shift can potentially enhance the reliability and safety of critical plant components while optimizing maintenance efforts and minimizing unnecessary outages. The NRC is continuing to explore the regulatory aspects of advanced technologies as part of inservice inspection and inservice testing (ISI and IST) programs by pursuing additional research in this technical area.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Influence of americium complexation on the radiation-induced chemical reactivity of sulfophenyl bistriazinyl pyridine (SO 3 -Ph-BTP) towards the nitrate radical

Sulfophenyl bistriazinyl pyridine (SO 3 -Ph-BTP) is a hydrophilic organic ligand used to separate actinides and lanthanides. Dose accumulation and time-resolved studies have previously provided insight into the radiolytic stability and degradation pathways of SO 3 -Ph-BTP in reprocessing environments, but no study has yet addressed the impact of minor actinide complexation on the radiation chemistry of this ligand. To begin to fill this knowledge gap, a systematic, time-resolved study exploring the reactivity of the nitrate radical (NO 3 • ) with SO 3 -Ph-BTP in the presence of trivalent americium, Am(III), has been conducted, which demonstrates enhanced reactivity (an order of magnitude faster) upon metal complexation.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2025

A mathematical approach to using the forgetting curve to evaluate experience and training factors in human reliability analysis

Traditional human reliability analysis (HRA) methods have difficulty dealing with the dynamic nature of factors such as time and rely on static and expert-judgment-based assessments of performance-shaping factors (PSFs) across limited levels. In this study, we introduce a mathematical approach for dynamically evaluating the experience and training PSF. Our proposed method integrates the psychological concept of the “forgetting curve” to evaluate how PSFs are impacted by the number of trainings and the time elapsed since training. To confirm the validity of the model, we provide experimental data fitted by identifying the quantitative relationship between training and human performance. This research enables dynamic and objective assessments, thus reducing reliance on subjective expert judgment and improving the accuracy of HRA.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

Monitoring Methods for Early Detection of Inadvertent Fission Product Release at the Advanced Test Reactor

Isotope effluent data obtained during three instances of experiment failures at the Advanced Test Reactor (ATR) are analyzed to provide an overview of the methods used to detect initial signs of unintended fission product release. The data is contextualized with the operational experience, including means of identification and subsequent mitigation strategies, gained during these events. General trends as well as variations in isotopic behavior between the three failures are explored. Background on the Real Time Monitor, a High Purity Germanium detector, and other fission product monitoring systems utilized at the Advanced Test Reactor is also provided. The presented analysis was used to establish administrative action levels which are currently utilized by ATR for early detection of experiment fission product release. Early identification provides time to make programmatic decisions before approaching safety and environmental limits.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

Smart Planning for Radioactive Source Transport Advanced Tools for Increased Safety and Efficiency

End-of-life (EOL) management of high-activity radioactive sources is made uniquely challenging by the inherent risks associated with storage and transportation of these sources, the complex logistics involved, and the strict requirements for regulatory compliance. Traditional methods lack comprehensive tools for accurate site assessments and precision planning for the transportation of radioactive sources. They also frequently fail to provide the adaptability required to consider diverse operational environments, resulting in inefficiencies and potential safety concerns. This paper introduces a novel software solution developed to address these issues by integrating advanced technologies such as light detection and ranging (LiDAR)-based 3D environment modeling, smart dynamic route planning, and customizable measurement functionalities. This software enables detailed terrain visualizations, facilitating thorough environmental assessments and enabling users to virtually navigate, analyze, and plan site-specific operations. Among the key features are a user-centric interface for virtual navigation, precise site measurement tools for site evaluations, interactive visualizations that highlight potential operational hazards, dynamic route planning capabilities, and real-time collision detection to promote safe workflows. By demonstrating the effectiveness of this tool through real-world application, the present work underscores the tool’s potential to revolutionize radioactive source EOL management by improving operational efficiencies, minimizing risk, and advancing the state of practice to achieve suitable and secure radioactive material handling.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Successes and Challenges of the Integrated Waste Treatment Unit's 50-Day Demonstration Run - 20151

The Integrated Waste Treatment Unit (IWTU) has been tasked with using steam reforming technology to treat 3.4 million liters (900,000 gallons) of radioactive sodium-bearing waste (SBW) at the Idaho National Laboratory (INL). The project has been hampered by a number of processing issues associated with process equipment during simulant testing, with the two main issues involving the fluidizing gas distribution system in the main reactor vessel, the denitration mineralization reformer (DMR), and high differential pressures in the main filter vessel, the process gas filter (PGF). Following the replacement of the fluidizing gas distribution system in the DMR, a successful 30-day demonstration run was performed that revealed unacceptable bridging in the filter bundles of the PGF. The bundles were replaced with new ones, and a 50-day demonstration run was performed. The main purpose of this run was to operate the IWTU over a range of operating conditions. Overall, the run was a success; however, there were multiple challenges that arose from the run. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Large Eddy Simulation of Low-Reynolds-Number Turbulent Flow of Low-Prandtl-Number Fluid in a Tight Lattice Bundle for Assessment of Reynolds-Averaged Navier-Stokes Turbulence Model

The MARVEL (Microreactor Applications Research Validation and Evaluation) microreactor utilizes natural circulation as core cooling mechanism and liquid metal as a primary coolant. Moreover, the reactor core has a pitch-to-diameter ratio of 1.056, which is considered a tight lattice configuration. Numerous studies have widely reported that Reynolds-Averaged Navier-Stokes (RANS) turbulence models inaccurately predict heat transfer in liquid metals and fail to capture flow pulsations that can occur within tight lattices, leading to further inaccuracies in simulation results. Therefore, evaluating the accuracy of RANS turbulence models in the thermal-hydraulic analysis of the MARVEL microreactor core is crucial for assessing reactor safety. In this study, a Large Eddy Simulation (LES) of the MARVEL microreactor core subchannel was conducted and compared with a RANS simulation to evaluate the accuracies and conservatism of the RANS model. The flow pulsation in a tight lattice predicted by LES enhanced the heat transfer, whereas the RANS model underpredicted it. Consequently, the RANS model predicted the peak cladding temperature higher than the LES model, but the discrepancy between the two approaches was not significant due to the good thermal characteristics of the liquid metal. It can be concluded that the steady-state RANS model is effective for the thermal analysis of liquid-metal-cooled MARVEL microreactor core and can provide conservative predictions from a safety analysis perspective.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

Flammable Gas Generation and Control at the Idaho Completion Project Legacy Combustible Gas Generation - 20191

On April 11, 2018, four drums containing transuranic waste at the Idaho National Laboratory underwent over-pressurization, ejecting their lids and spreading radiological waste within a facility. An investigation has found that waste in the drums generated methane gas, which contributed to the event. Subsequent to the investigation, the potential for drums to have methane and other flammable analytes whose concentrations could approach or exceed the lower flammability limit (LFL) and the adequacy of the controls to prevent or mitigate a possible deflagration was evaluated. An extensive review of the historical records was performed to determine how many drums exceed the LFL. The historical record identified a small quantity of drums that exceeded the LFL for xylene, hydrogen, and methane. The primary codified applicable code or standard for handling drums is 29 CFR 1910.120 (j) 'Handling Drums and Containers.' It is used throughout the commercial and government sector. This code is integrated throughout the Fluor Idaho safety management programs and procedures to control the Standard Industrial Hazards (SIHs) associated with drum and container handling. Example requirements include: - Drums and containers are inspected, and their integrity assured prior to being moved. - Site operations are organized to minimize the amount of drum or container movement. - When there is a reasonable possibility of flammable atmospheres being present, material handling equipment and hand tools are of the type to prevent sources of ignition. Drums and containers under pressure, as evidenced by bulging or swelling, are not moved until the cause for excess pressure is determined and appropriate containment procedures have been implemented to protect employees from explosive relief of the drum. Drums which exceed the WIPP FGA limit have NCRs associated with them. Once a drum fails flammable gas sampling, an NCR is initiated. The NCR drives placement into NCR dense pack rows. A standing order drives placement of the FGA failures to a single planer segregation location. These drums are not allowed for further processing or shipment until the NCR has been cleared. Once in the single planer rows, follow-up FGA testing is performed by CCP. The NCR is dispositioned by follow-up sampling occurring after adequate time has been provided for the gas to diffuse through the filter assembly, or, in some cases, the drums are repackaged into additional drums to reduce the source term. To date, the ICP project has performed approximately 1.5 million drums moves without a deflagration event due to combustible gas generation. Flammable gas generation in a transuranic waste drum is not unique to the ICP but is common across the DOE complex. Based on the experience at Idaho, application of industry standards is sufficient to control the risk of drum deflagration due to drum movements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Enabling the Write-Once Read-Many (WORM) Dashboard and Web Interface for HPC Data Storage

Idaho National Laboratory (INL), supported by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Nuclear Science User Facilities, provides access to supercomputer systems and data storage along with support staff for system management, software installation, cybersecurity, and user support to the broader DOE-NE user community. A key component of this is providing long-term or perpetual data storage as part of efforts to improve reproducibility in modeling and simulation in addition to supporting simulations with regulatory compliance requirements. This milestone details the deployment of a new type of storage option for users requiring tamper resistant perpetual storage of data and the associated support for data control and release via the data storage web interface. The perpetual data storage system is integrated within the Open OnDemand framework deployed at INL and is called the ?WORM? in recognition of its write-once read-many characteristics.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Review of recent activities with MOOSE, an open-source finite element & finite volume multi-fidelity simulation framework

Modeling and simulation are an increasing part of engineering. This is undoubtedly driven by the high costs of constructing experimental facilities, but also enabled by the exponential increase in computing powers over the last decades, which allows computational models to be closer than ever to reality. One of the main drivers for the development of MOOSE is supporting advanced nuclear reactor simulations. A challenging aspect of modeling advanced nuclear reactors is the plurality of physics involved, including neutronics, thermal hydraulics and fuel performance. These physics are all coupled to some extent and are generally solved in a sequential but iterative fashion. The United States (U.S.) national laboratories have been developing MOOSE, an open source multiphysics framework since its inception at the Idaho National Laboratory (INL) in 2008. This framework enables seamless coupling of multiphysics simulations and facilitates the implementation of new physics and material governing laws. It is continuously expanded with novel numerical methods and new pre-implemented physics module. Numerous applications, developed within the Department of Energy (DOE) laboratories, academia, and industry, including outside of nuclear engineering, have been developed to study specialized physics problems. International collaborations are welcome on this open-source modeling and simulation project.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Nuclear—thermal energy storage configurations for industrial combined heat and power supply—conceptual and thermodynamic study with high temperature gas-cooled reactor

Nuclear systems are promising candidates for delivering resilient heat and power for future energy security and independence. Traditionally, nuclear plants have been used for baseload electricity production and cogeneration of heat has seen relatively limited application utilizing typically only small portion of a reactor's thermal output. This paradigm may shift due to the increasing penetration of intermittent renewables and need for resource flexibility, various decarbonization efforts aimed at both electricity and heat demands, along with the perspective of small modular nuclear reactor applications, which can be sized based on local industrial needs. Here, this study provides a comprehensive guide for the nuclear and industrial sectors, emphasizing controllability in the combined heat and power configuration options for high temperature gas-cooled reactor and process steam supply. It investigates the integration of thermal energy storage to improve nuclear energy's responsiveness to varying industrial demands. The study emphasizes placing thermal energy storage between the nuclear primary loop and steam cycle to achieve greater efficiency and flexibility in power and heat output, surpassing traditional combined heat and power systems and avoiding efficiency losses seen in other thermal energy storage integration approaches.

combined heat and power (CHP)↗
DOE OSTI2026

Hazard and risk analysis framework for nuclear power plant–based integrated energy systems

Employing integrated energy systems (IESs) with nuclear power plants (NPPs) can improve NPP utilization by leveraging dedicated thermal and electric power delivery, but it may also increase operational safety risks. This paper presents a framework to identify and quantify hazards and risks for such IESs. The framework combines accidentology to review past industrial accidents with failure modes and effects analysis (FMEA) to identify potential future incidents. Hydrogen explosion and toxic chemical release hazards are of particular concern. Explosion consequences are quantified using the Bauwens-Dorofeev (Bauwens) and trinitrotoluene equivalent mass (TNT-EM) methods, while chemical release consequences are computed using the Gaussian atmospheric dispersion method. Operational disturbances from direct electrical and thermal integration that may affect NPP safety are modeled using probabilistic risk analysis (PRA). Hazards and risks are then evaluated for regulatory compliance. The framework is applied to IESs comprising pressurized or boiling water reactors supplying three levels of thermal and electrical power to industrial customers. Case studies include high-temperature steam electrolysis hydrogen plants of varying capacities and a synthetic fuel production plant. Sensitivity analysis examines piping component failures in the PRA model as a precursor to cost estimation for thermal extraction line design. Additionally, Fussel-Vessely (FV) and risk increase importance (RII) measures identify risk-informed design improvements for the thermal extraction system. FMEA highlights hazards such as loss of offsite power, prompt loss of electrical load, loss of thermal output, and immediate steam diversion, in addition to hydrogen explosions and toxic chemical releases. Both Bauwens and TNT-EM methods suggest maintaining several hundred meters of separation between the NPP and hydrogen facility to mitigate explosion risks. PRA results show a maximum initiating event frequency increase of 1.15% and an overall risk increase of 0.28%. Importance measure analysis identifies upstream pipe leak isolation components as critical. Evaluating the results against safety regulations, it is concluded that hazards and risks can be managed to comply with regulations through risk-informed thermal and electrical connection designs, component selection, maintenance programs, and safe separation distances between NPPs and integrated industrial facilities.

08 - HYDROGEN↗