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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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Thermal-Fluid Analysis of a Liquid-Cooled Battery Module for Electrified Aircraft

The development of safe, energy-dense batteries is critical to advancing hybrid electric and fully electrified aircraft propulsion. Achieving this capability requires a thermal management system that can maintain battery performance and safety under demanding operational conditions. The objective of this project is to support the maturation of next-generation lithium-ion batteries for electrified aircraft by conducting performance testing on integrated battery modules, specifically a 2-cell series configuration module housed within an aluminum enclosure. Designed to operate at a nominal 7.2 V with discharge rates up to 2.5C, the module will eventually be used to power an electric motor and DC-DC converter, generating substantial thermal loads that must be effectively managed to increase the usable energy and power density of electrified aircraft. To address these thermal challenges, this study presents the development and thermal-fluid analysis of a liquid-cooled thermal management system. While the full aircraft architecture utilizes an eight-string configuration, the present work evaluates a representative single-string water coolant loop to characterize baseline performance. The active cooling loop circulates water through a reservoir, pump, the battery module, and a variable area flow meter. Key performance metrics including component-level temperatures, mass flow rates, and pressure drops are quantified across the loop. Across discharge rates ranging from 0.5C to 2.5C, the active thermal loop consistently and effectively removed heat from the module, validating the design approach and confirming readiness for further development. The validated thermal performance indicates a path toward scalable battery modules that could enable energy and power dense systems for hybrid electric aircraft.

Electrified Aircraft↗
DOE OSTI2022

Designing Secure and Resilient Cyber-Physical Systems Using Formal Models

This work-in-progress paper proposes a design methodology that addresses the complexity and heterogeneity of cyber-physical systems (CPS) while simultaneously proving resilient control logic and security properties. The design methodology involves a formal methods-based approach by translating the complex control logic and security properties of a water flow CPS into timed automata. Timed automata are a formal model that describes system behaviors and properties using mathematics-based logic languages with precision. Due to the semantics that are used in developing the formal models, verification techniques, such as theorem proving and model checking, are used to mathematically prove the specifications and security properties of the CPS. This work-in-progress paper aims to highlight the need for formalizing plant models by creating a timed automata of the physical portions of the water flow CPS. Extending the time automata with control logic, network security, and privacy control processes is investigated. The final model will be formally verified to prove the design specifications of the water flow CPS to ensure efficacy and security.

42 ENGINEERING↗
DOE OSTI2016

Software Dependability for Digital Instrumentation and Control Systems in Nuclear Power Plants - Lessons Learnt from a DOE Sponsored Project

This paper discusses the lessons learnt from a sponsored study undertaken under the sponsorship of the Department of Energy by a team of scientists at the Ohio State University in collaboration with Technology Resources and experts around the world. This paper makes recommendations for future research and studies based on these lessons learnt. The purpose of the research was to provide a software dependability quantification framework that would allow its evaluation along the software development life cycle for safety critical instrumentation and control systems (I and C) such as those found in nuclear power plants. A typical example of safety critical system is the reactor protection system. The research funded by the Department of Energy built on preliminary research carried out for the US NRC that had focused solely on software reliability quantification. Quantification of software dependability is important for safety critical I and C because the regulatory framework in use in the United States for nuclear power plants is risk informed. Quantification of risk requires quantification of all contributors to risk and those include instrumentation and control systems. The purpose of the study was to help developers of software-based I and C with an early evaluation of the potential risk an application would represent at different stages of the software development life cycle by building gates at the end of each life cycle phase. Life cycle phases of software development comprehend phases such as requirements, design, coding and test. The gates' function would be to stop development from moving forward if the assessed dependability objective for that phase were not met. Gates thresholds were to be defined based on measurements collected during the life cycle phase under evaluation and phases that preceded it and an understanding of the mechanisms that would lead to threatening the dependability objectives. The information contained in the gates would in addition help steer the development towards areas that needed improvement. The following are four high level conclusions to our study. First the study demonstrates a lack of consensus on mechanisms that will threaten the dependability attributes. This lack of consensus should be resolved. The lack of consensus appears for all measures to the exception of reliability. Second, measures should be developed for software availability and safety. Third, for all dependability attributes (with the exception of reliability) consistency of the measurement threads should evaluated and measurement-based development gates derived. Finally our ability to perform such type of research depends in a significant manner on access to vendor and/or utility data. The licensing environment and the first-of-a-kind nature of current software-based applications for nuclear I and C make it difficult to collect the relevant data. An approach which would either increase access to data by providing 'immunity' to the vendor or utility or the development of a software-based I and C for the sole purpose of data collection for research should be investigated. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Fiber Optic Sensor for Corrosion Monitoring in Molten Salt Irradiation Experiments

Molten salt reactors (MSRs) are one of the more promising advanced reactor concepts because they offer passive safety, high-temperature process heat, and the potential for improved fuel cycle economics through online reprocessing [1]. However, before MSRs can be reliably deployed, structural materials must be identified that can survive extended exposure to highly-corrosive salts during irradiation at high temperatures. The U.S. Department of Energy (DOE) is funding the design and eventual construction of a Versatile Test Reactor (VTR) that will include experimental facilities for testing under conditions representative of several advanced reactors, including MSRs [2]. These facilities will allow testing of candidate structural materials during irradiation with flowing salt. If properly designed, these experiments will allow for post-irradiation characterization of weight loss in the structural material coupons. However, significantly more information could be gained if the corrosion could be monitored in situ.There are currently no commercially available sensors for in situ corrosion monitoring at temperatures in the range of 500–900°C [3]. An in situ corrosion sensor would also have to survive high-dose neutron irradiation and extended exposure chemically aggressive media (e.g., molten salts). While there are ongoing efforts to use the change in magnetic susceptibility of Cr-containing alloys, this technique is limited to Cr-containing structural materials and there are concerns when using a magnetic susceptibility sensor at temperatures approaching the Curie temperature of iron (770°C) [4].

Petrie, Christian M.↗
DOE OSTI2025

Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model Updates

This work presents the latest improvements to, and investigations performed with, the Fluoride-Cooled High-Temperature Pebble-Bed Reactor reference plant models for the United States Nuclear Regulatory Commission. These models, developed with the Comprehensive Reactor Analysis Bundle, or BlueCRAB, serve as the foundation for the future development of detailed design evaluation models based on license applications. BlueCRAB is the code suite proposed for non-light-water reactor systems safety analyses, and it incorporates various simulation tools developed by the Nuclear Energy Advance Modeling and Simulation program, including the Griffin code for reactor physics, the Pronghorn and SAM codes for core thermal fluids, the BISON code for solid conduction and fuel performance, and the SAM code for system analysis. The primary objective of this work is to assess the level of readiness of BlueCRAB for modeling fluoride-cooled high-temperature pebble-bed reactors. We first developed numerical models in BlueCRAB that include the key physics for this technology, ensuring an adequate level of fidelity for modeling the core performance during accident scenarios. This was followed by simulation of transient scenarios, two loss-of-forced-cooling events (one protected and one unprotected), and two control rod withdrawal events (one delayed and one prompt supercritical reactivity insertion). The analysis includes comparisons between the 2-D thermal fluid porous media models in Pronghorn and SAM, comparisons between coupled Pronghorn-Griffin and coupled SAM-Griffin models for two loss-of-forced cooling events and one control rod withdrawal event, and comparisons between SAM single-solve and domain-overlapping approaches for multi-scale thermal fluid coupling. In addition, we performed comparisons between 3-D, 2-D, and 0-D neutronic models for the two control rod withdrawal scenarios with Pronghorn-Griffin. The results show that the BlueCRAB models led to physically intuitive solutions for the scenarios examined. The changes in the various scalar and vector fields such as the neutron flux, power, temperatures, densities, pressures, and velocities are all within the expected ranges, and their distributions can be explained from the system response of the transients and the geometric and material variations. Several comparisons suggest that the porous media models in Pronghorn and SAM can lead to similar solutions, even though they are based on different methodologies. The simulations demonstrate that there are differences between the various levels of fidelity, and it is advisable to have flexible tools that can cover the breadth and depth of needs that may arise in future technical evaluations. We believe that BlueCRAB’s capabilities represent a significant asset for confirmatory analyses aimed at resolving important safety questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2008

NASA's Ares I and Ares V Launch Vehicles -- Effective Space Operations Through Efficient Ground Operations

The United States (U.S.) plans to return to the Moon by 2020, with the development of a new human-rated space transportation system to replace the Space Shuttle, which is due for retirement in 2010 after it completes its missions of building the International Space Station and servicing the Hubble Space Telescope. Powering the future of space-based scientific exploration will be the Ares I Crew Launch Vehicle, which will transport the Orion Crew Exploration Vehicle to orbit where it will rendezvous with the Lunar Lander. which will be delivered by the Ares V Cargo Launch Vehicle. This new transportation infrastructure, developed by the National Aeronautics and Space Administration (NASA), will allow astronauts to leave low-Earth orbit for extended lunar exploration and preparation for the first footprint on Mars. All space-based operations begin and are controlled from Earth. NASA's philosophy is to deliver safe, reliable, and cost-effective solutions to sustain a multi-billion-dollar program across several decades. Leveraging 50 years of lessons learned, NASA is partnering with private industry, while building on proven hardware experience. This paper will discuss how the Engineering Directorate at NASA's Marshall Space Flight Center is working with the Ares Projects Office to streamline ground operations concepts and reduce costs. Currently, NASA's budget is around $17 billion, which is less than 1 percent of the U.S. Federal budget. Of this amount, NASA invests approximately $4.5 billion each year in Space Shuttle operations, regardless of whether the spacecraft is flying or not. The affordability requirement is for the Ares I to reduce this expense by 50 percent, in order to allow NASA to invest more in space-based scientific operations. Focusing on this metric, the Engineering Directorate provides several solutions-oriented approaches, including Lean/Six Sigma practices and streamlined hardware testing and integration, such as assembling major hardware elements before shipping to the Kennedy Space Center for launch operations. This paper provides top-level details for several cost saving initiatives, including both process and product improvements that will result in space transportation systems that are designed with operations efficiencies in mind. The Engineering Directorate provides both the intellectual capital embodied in an experienced workforce and unique facilities in which to validate the information technology tools that allow a nationwide team to collaboratively connect across miles that separate them and the engineering disciplines that integrate various piece parts into a whole system. As NASA transforms ground-based operations, it also is transitioning its workforce from an era of intense hands-on labor to a new one of mechanized conveniences and robust hardware with simpler interfaces. Ensuring that space exploration is on sound footing requires that operations efficiencies be designed into the transportation system and implemented in the development stage. Applying experience gained through decades of ground and space op'erations, while using value-added processes and modern business and engineering tools, is the philosophy upon which a new era of exploration will be built to solve some of the most pressing exploration challenges today -- namely, safety, reliability, and affordability.

Dumbacher, Daniel L.↗
NASA NTRS2011

Ares I-X Malfunction Turn Range Safety Analysis

Ares I-X was the designation given to the flight test version of the Ares I rocket which was developed by NASA (also known as the Crew Launch Vehicle (CLV) component of the Constellation Program). The Ares I-X flight test vehicle achieved a successful flight test on October 28, 2009, from Pad LC-39B at Kennedy Space Center, Florida (KSC). As part of the flight plan approval for the test vehicle, a range safety malfunction turn analysis was performed to support the risk assessment and vehicle destruct criteria development processes. Several vehicle failure scenarios were identified which could have caused the vehicle trajectory to deviate from its normal flight path. The effects of these failures were evaluated with an Ares I-X 6 degrees-of-freedom (6-DOF) digital simulation, using the Program to Optimize Simulated Trajectories Version II (POST2) simulation tool. The Ares I-X simulation analysis provided output files containing vehicle trajectory state information. These were used by other risk assessment and vehicle debris trajectory simulation tools to determine the risk to personnel and facilities in the vicinity of the launch area at KSC, and to develop the vehicle destruct criteria used by the flight test range safety officer in the event of a flight test anomaly of the vehicle. The simulation analysis approach used for this study is described, including descriptions of the failure modes which were considered and the underlying assumptions and ground rules of the study.

Beaty, J. R.↗
DOE OSTI2019

Computational study of full-scale VHTR lower plenum for turbulent mixing assessment

Next-generation nuclear reactors are poised to efficiently provide reliable power at enhanced safety levels for many years to come. Among the fundamental designs for these reactors is the very-hightemperature reactor (VHTR), which employs helium as the primary coolant and has the potential of reaching elevated temperatures such that chemical processing (e.g., electrolysis for hydrogen production) is viable alongside traditional electricity generation. This and other next-generation reactors also represent a strategic bridge for ultimately transitioning from fossil fuel-dominated energy portfolios to one where renewable options abound. For the VHTR, the ultimate goal of constructing a new plant will be preceded by additional fundamental research aimed at generating trusted models for behavior prediction under normal and accident scenarios. This paper presents a detailed computational fluid dynamics simulation of the lower plenum, where hot coolant from the core mixes together in a turbulent fashion before traveling to power conversion equipment. Because the flow is expected to enter the lower plenum across a wide range of temperatures and velocities, concerns exist when the mixing is incomplete. The potential hot spots on cylindrical support pedestals and the uniformity of the main outlet are the primary metrics of interest in this study, along with flow velocities and temperatures at different locations within the lower plenum. Three locations are probed in detail and suggest a large range exists for approach velocities and temperatures. A large degree of stratification is also seen on the surfaces of the support pedestals, suggesting a facility should accommodate testing for such behavior. The characterizations presented provide the valuable data needed in order to design appropriate experimental testbeds, where scaled modeling can be carried out in a manner meaningful in predicting the full-scale behavior.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Artificial Intelligence in Nuclear Safeguards; Evaluating Safeguards and Security Risks and Benefits for Advanced and Small Modular Reactor Deployments

Rapidly growing interest in advanced and small modular reactor (A/SMR) technologies presents challenges as well as opportunities for implementing international safeguards and security. A/SMR deployments are expected to be more numerous, more geographically dispersed, and more varied in their designs, placing new demands on the data systems and analytical tools used to support oversight (Alberti et al., 2023; Canadian Nuclear Safety Commission et al., 2024). Because of this variability, the importance and reliance on data systems for A/SMR deployments is expected to be higher than for previous reactor generations. Artificial Intelligence and Machine Learning (AI/ML) offer potential capabilities to address the high variability inherent in A/SMR technology. The beneficiaries of AI-assisted tools include facility operators, government regulators, IAEA inspectors, and A/SMR vendors. This report analyzes how AI/ML-assisted technologies can strengthen the implementation of IAEA safeguards and security measures. It also identifies AI-assisted tools to strengthen operator, facility, and regulator knowledge management practices and examines the potential risks AI/ML-based tools may introduce to IAEA safeguards and security efforts. It concludes with a set of hypothetical, standards-style requirements for AI/ML systems used in safeguards contexts, grounded in an inspector-centric view of system verification. Despite the potential benefits of AI/ML systems, understanding potential intentional and unintentional failure modes is critical for ensuring adequate protection of nuclear materials and facilities. Unique features of A/SMRs including sealed cores, remote and novel paradigms of operation, off-site reactor fabrication, novel fuel forms, and varied refueling requirements, introduce challenges for traditional safeguards technological approaches (Pensado et al., 2024; Federation of American Scientists, 2025). AI/ML systems deployed to address these challenges may introduce new risks requiring systematic evaluation rooted in both AI-specific risk frameworks, such as the NIST AI Risk Management Framework (NIST AI RMF), and established cyber risk management standards such as NIST SP 800-30 (National Institute of Standards and Technology [NIST], 2023; NIST, 2012).

97 MATHEMATICS AND COMPUTING↗
NASA NTRS2018

Optimization of Second Fault Detection Thresholds to Maximize Mission POS

In order to support manned spaceflight safety requirements, the Space Launch System (SLS) has defined program-level requirements for key systems to ensure successful operation under single fault conditions. To accommodate this with regards to Navigation, the SLS utilizes an internally redundant Inertial Navigation System (INS) with built-in capability to detect, isolate, and recover from first failure conditions and still maintain adherence to performance requirements. The unit utilizes multiple hardware- and software-level techniques to enable detection, isolation, and recovery from these events in terms of its built-in Fault Detection, Isolation, and Recovery (FDIR) algorithms. Successful operation is defined in terms of sufficient navigation accuracy at insertion while operating under worst case single sensor outages (gyroscope and accelerometer faults at launch). In addition to first fault detection and recovery, the SLS program has also levied requirements relating to the capability of the INS to detect a second fault, tracking any unacceptable uncertainty in knowledge of the vehicle's state. This detection functionality is required in order to feed abort analysis and ensure crew safety. Increases in navigation state error and sensor faults can drive the vehicle outside of its operational as-designed environments and outside of its performance envelope causing loss of mission, or worse, loss of crew. The criteria for operation under second faults allows for a larger set of achievable missions in terms of potential fault conditions, due to the INS operating at the edge of its capability. As this performance is defined and controlled at the vehicle level, it allows for the use of system level margins to increase probability of mission success on the operational edges of the design space. Due to the implications of the vehicle response to abort conditions (such as a potentially failed INS), it is important to consider a wide range of failure scenarios in terms of both magnitude and time. As such, the Navigation team is taking advantage of the INS's capability to schedule and change fault detection thresholds in flight. These values are optimized along a nominal trajectory in order to maximize probability of mission success, and reducing the probability of false positives (defined as when the INS would report a second fault condition resulting in loss of mission, but the vehicle would still meet insertion requirements within system-level margins). This paper will describe an optimization approach using Genetic Algorithms to tune the threshold parameters to maximize vehicle resilience to second fault events as a function of potential fault magnitude and time of fault over an ascent mission profile. The analysis approach, and performance assessment of the results will be presented to demonstrate the applicability of this process to second fault detection to maximize mission probability of success.

Anzalone, Evan↗
DOE OSTI2025

Openpronghorn

OpenPronghorn is a simulation tool specifically tailored for modeling thermal-hydraulic phenomena in advanced nuclear reactors. It is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE), an open-source platform that facilitates the development of high-performance scientific computing applications. OpenPronghorn solves the Navier-Stokes equations, which describe the conservation of mass, momentum, and energy in fluid flows, using the finite volume numerical method. The code supports a wide range of fluid flow conditions that are applicable to nuclear reactors, including incompressible and weakly compressible flows, as well as single-phase and multiphase flows. It is capable of modeling diverse flow regimes, including laminar and turbulent flows, using various turbulence models such as the standard k-epsilon models, the v2f model, and the mixing length model. For multiphase flows, OpenPronghorn employs a mixture a Eulerian modeling approach with mixture, drift-flux, and full Eulerian models, and includes open-sourced interfacial transfer correlations for drag, exchange, and heat transfer coming from the scientific literature. OpenPronghorn's modular design allows it to handle multiscale simulations, ranging from detailed Reynolds-Averaged Navier Stokes (RANS) simulations to coarse-mesh and lumped parameter models. This flexibility enables users to perform high-fidelity simulations of specific reactor components as well as system-level analyses of entire reactor circuits. The code can be coupled with other MOOSE-based tools using the MultiApp system, allowing for the transfer of coupling quantities such as mass flow rates, heat fluxes, and boundary conditions between different simulation scales. One of the main features of OpenPronghorn is the it includes built-in validation cases from the open-source scientific literature and supports the implementation of user-defined models and correlations through MOOSE's FunctorMaterial system. OpenPronghorn is designed to be computationally efficient, leveraging the SIMPLE projection method for large-scale problems, and can be run on high-performance computing systems to handle the extensive computational demands of detailed reactor simulations. Overall, OpenPronghorn is a versatile and robust tool that provides critical insights into the thermal-hydraulic behavior of advanced nuclear reactors, supporting the design, safety, and optimization of next-generation nuclear energy systems.

Retamales, Mauricio Eduardo Tano [Idaho National L↗
DOE OSTI2023

Experiences of Member States in Building a Nuclear Security Infrastructure for New Nuclear Power Programmes

A nuclear power programme has many benefits, but it is a major undertaking that requires careful planning and preparation as well as a substantial investment in time. It also necessitates the establishment of a sustainable national infrastructure with committed and sustained financial and human resources. While nuclear power is not unique in this respect, it is different from other sources of energy because of the risks associated with the use of nuclear material, as well as the need for compliance with international legal instruments, internationally accepted nuclear safety standards, nuclear security guidance, and safeguards requirements. The IAEA’s Milestones in the Development of a National Infrastructure for Nuclear Power, IAEA Nuclear Energy Series No. NG-G-3.1 (Rev. 1), defines a phased approach that identifies 19 infrastructure issues that should be addressed in each of the three phases of the development of a nuclear infrastructure for a nuclear power programme. The publication Evaluation of the Status of National Nuclear Infrastructure Development IAEA Nuclear Energy Series NG-T-3.2 (Rev. 2) provides a methodology to determine the status of the infrastructure conditions covering all 19 issues identified in the Milestones Approach. A companion IAEA Implementing Guide, Establishing the Nuclear Security Infrastructure for a Nuclear Power Programme, IAEA Nuclear Security Series No. 19, provides guidance on the recommended actions to be taken by a State to establish an effective national nuclear security infrastructure for a nuclear power programme using the Milestones Approach. This publication is intended to provide guidance based on the experiences and good practices of Member States with embarking nuclear power programmes as well as Member States that are expanding their nuclear power programmes. The guidance is in the form of case studies, which highlight challenges, issues, and solutions identified by Member States as lessons learned for new nuclear power programmes. This TECDOC is designed to present the experiences of Member States to assist other Member States in developing a nuclear security programme. It addresses the key actions necessary to establish an effective national nuclear security infrastructure for a nuclear power programme consistent with the three phases of the IAEA Milestones Approach. The Member States that provided case studies in this document are at different stages in the development of a nuclear power programme.

42 ENGINEERING↗
NASA NTRS1995

A Comprehensive Approach to Management of Workplace and Environmental Noise at NASA Lewis Research Center

NASA Lewis Research Center is home to more than 100 experimental research testing facilities and laboratories, including large wind tunnels and engine test cells, which in combination create a varied and complex noise environment. Much of the equipment was manufactured prior to the enactment of legislation limiting product noise emissions or occupational noise exposure. Routine facility maintenance and associated construction also contributes to a noise exposure management responsibility which is equal in magnitude and scope to that of several small industrial companies. The Noise Program, centrally managed within the Office of Environmental Programs at LRC, maintains overall responsibility for hearing conservation, community noise control, and acoustical and noise control engineering. Centralized management of the LRC Noise Program facilitates the timely development and implementation of engineered noise control solutions for problems identified via either the Hearing Conservation of Community Noise Program. The key element of the Lewis Research Center Noise Program, Acoustical and Noise Control Engineering Services, is focused on developing solutions that permanently reduce employee and community noise exposure and maximize research productivity by reducing or eliminating administrative and operational controls and by improving the safety and comfort of the work environment. The Hearing Conservation Program provides noise exposure assessment, medical monitoring, and training for civil servant and contractor employees. The Community Noise Program aims to maintain the support of LRC's neighboring communities while enabling necessary research operations to accomplish their programmatic goals. Noise control engineering capability resides within the Noise Program. The noise control engineering, based on specific exposure limits, is a fundamental consideration throughout the design phase of new test facilities, labs, and office buildings. In summary, the Noise Program addresses hearing conservation, community noise control, and acoustical and noise control engineering.

Cooper, Beth A.↗
NASA NTRS2000

Shuttle Upgrade Using 5-Segment Booster (FSB)

In support of NASA's continuing effort to improve the over-all safety and reliability of the Shuttle system- a 5-segment booster (FSB) has been identified as an approach to satisfy that overall objective. To assess the feasibility of a 5-segment booster approach, NASA issued a feasibility study contract to evaluate the potential of a 5-segment booster to improve the overall capability of the Shuttle system, especially evaluating the potential to increase the system reliability and safety. In order to effectively evaluate the feasibility of the 5-segment concept, a four-member contractor team was established under the direction of NASA Marshall Space Flight Center (MSFC). MSFC provided the overall program oversight and integration as well as program contractual management. The contractor team consisted of Thiokol, Boeing North American Huntington Beach (BNA), Lockheed Martin Michoud Space Systems (LMMSS) and United Space Alliance (USA) and their subcontractor bd Systems (Control Dynamics Division, Huntsville, AL). United Space Alliance included the former members of United Space Booster Incorporated (USBI) who managed the booster element portion of the current Shuttle solid rocket boosters. Thiokol was responsible for the overall integration and coordination of the contractor team across all of the booster elements. They were also responsible for all of the motor modification evaluations. Boeing North American (BNA) was responsible for all systems integration analyses, generation of loads and environments. and performance and abort mode capabilities. Lockheed Martin Michoud Space Systems (LMMSS) was responsible for evaluating the impacts of any changes to the booster on the external tank (ET), and evaluating any design changes on the external tank necessary to accommodate the FSB. USA. including the former USBI contingent. was responsible for evaluating any modifications to facilities at the launch site as well as any booster component design modifications.

Sauvageau, Donald R.↗
NASA NTRS2023

Establishing Standardized Test Methods for Evaluating Space Suit Gloves

The Artemis space suit glove environmental protection garment (EPG) will be the first line of protection used to protect the crewmember’s hands from the environments encountered during extravehicular activity (EVA). As the Artemis missions will include more extreme environments, it is necessary to define tests to evaluate if gloves can meet new, challenging requirements. The NASA Engineering and Safety Center (NESC)has funded an effort with the goal to help vendors to produce a space suit glove that is sufficiently durable to meet the needs of the Artemis missions. Three problems were identified in meeting that goal. First, there are no standardized tests defined to evaluate the durability of space suit gloves for the extreme lunar environments. Second, there is insufficient data on glove performance in a lunar environment from which to compare new designs. And third, current ISS glove TMG fabrics are unlikely to be sufficient to meet Lunar requirements. Four tasks were established to answer different aspects of these three problems. Task 1 focused on the characterization of potential glove fabrics using ASTM standardized test methods when possible. When one did not exist, new standardized test procedures were devised. Task 2 focused on developing a procedure for characterizing the thermal performance of a glove assembly. Task 3 focused on devising a standard procedure for characterizing the dust/abrasion resistance of a space suit glove. Task 4 focused on devising a standard procedure for characterizing the cut resistance of a space suit glove while at cryogenic temperatures. The new procedures created in Tasks 2, 3 and 4 were then validated by testing the Phase VI glove to collect baseline data. This paper will summarize the approach used to establish standardized test methods for each task and will expound on the ASTM process to publish the new standards.

spacesuit↗
DOE OSTI2020

Addressing Human and Organizational Factors in Nuclear Industry Modernization: An Operationally Focused Approach to Process and Methodology

Utility owners and operators of commercial nuclear power plants in the United States (U.S.) are and will be modernizing their nuclear power plants by performing a digital transformation involving design of an integrated set of systems that together enable a technology centric operating plant. The Plant Modernization Pathway of the U.S. Department of Energy Light Water Reactor Sustainability Program has a strategic action plan that lays the groundwork for a digital transformation of the nuclear industry. The model for this transformation is an advanced concept of operations, with an end point vision, “To achieve the maximum aggregate benefit enabled by this digital transformation.” To achieve this, the digital infrastructure for a nuclear plant must be designed as an integrated set of systems that together enable a technology centric operating model. The digital transformation process obviously needs to involve technology considerations and systems engineering, but it also needs to include human and organizational expertise. Thus, human and organizational factors, including sociotechnical systems methods and techniques (e.g., Cognitive Systems Engineering, Systems Theoretic Accident Modeling and Processes, human systems integration, and Macroergonomics) need to be considered for digital transformation projects in order to effectively integrate human and organizational expertise efforts into the new work system that results from nuclear power plant digital modernization. That is, the work system is the basic unit of sociotechnical systems analysis and contains three components: personnel, technical, and organization and management. These components should be jointly optimized with respect to the interdependence of systems performance criteria of effectiveness, efficiency and safety. Joint optimization can be achieved through the application of three human and organization functions: knowledge representation, knowledge elicitation, and cross-functional integration. This report provides a strategic framework for effective integration of human and organizational expertise within nuclear power plant digital modernization efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2024

Test and Analysis of the 8-foot Diameter Cylindrical Sandwich Composite Test Article CTA8.2B: As part of the NASA Engineering and Safety Center Shell Buckling Knockdown Factor Project

The buckling response of thin-walled cylindrical structures have long been shown to be sensitive to various imperfections. That is, imperfections related to load introduction, boundary conditions, material property variation, and geometric deviations can all contribute to experimentally obtained buckling loads being lower than the buckling loads predicted for perfect representations of the same thin-walled cylindrical structure. To account for the influence of geometry-based imperfections of a thin-walled cylindrical structure on buckling response, a design factor (also known as a buckling knockdown factor) is used during the design phase. Guidelines for buckling knockdown factors are readily available in NASA SP-8007, Buckling of Thin Walled Circular Cylinders, which was updated in 2020. Prior to this recent update, NASA SP-8007 had not been updated since 1969, and since that time, computational analysis, experimental testing, and manufacturing methods have significantly improved. The NASA Engineering and Safety Center (NESC) Shell Buckling Knockdown Factor Project (SBKF, NESC Assessment 07-010-E) had the goal of developing buckling design recommendations for select classes of metallic and composite launch-vehicle structures [3]. Specifically, the test and analysis results from the fourth SBKF composite test article SBKF-P3-CYL-CTA8.2B, which is referred to as CTA8.2B, are described in this paper. This test was the fourth in a series of four tests on sandwich composite cylinders that will be used to experimentally validate analysis methods, which in turn can be used to develop new analysis-based shell buckling design guidelines for typical sandwich composite launch vehicle cylindrical structures. Publications describing previous test validation with analysis are available in Refs. CTA8.2B was designed to occupy one corner of the design space with approximately equal bending stiffnesses in the axial and transverse directions and have a larger ratio of radius to effective shell thickness. Each test article had a unique design and was designed to occupy a specific region of the design space. CTA8.2B was an 8-foot diameter honeycomb-core sandwich cylinder that was fabricated and tested at the NASA Marshall Space Flight Center (MSFC). The primary objectives of this test were to interrogate the structural capability of a composite test article, and to verify the test article design and analysis approach for cylinders subjected to axial compression and combined axial and bending loads. First, descriptions of the test article design and test are given in Section 2.0, and modeling and analysis methods used in support of the test article design and testing activities are described briefly in Section 3.0. Then, select test results are presented and compared to predicted results in Section 4.0, and concluding remarks are presented in Section 5.0. Finally, drawings of CTA8.2B and references used to support the test and analysis are provided in the Appendix.

Nonlinear analysis↗
DOE OSTI2024

The Monolithic Heat Pipe Microreactor Reference Plant Model

This work introduces a reference plant model for a generic monolithic heat-pipe-cooled microreactor. The model will serve as a springboard to develop future evaluation models in the licensing process of similar microreactor designs at the U.S. Nuclear Regulatory Commission. This model has been developed with the Comprehensive Reactor Analysis Bundle and its specifications are based on open literature publications for the eVinci TM design. BlueCRAB is the U.S. Nu- clear Regulatory Commission non-light-water reactor analysis system based on the Multiphysics Object-Oriented Simulation Environment framework, which can couple the Griffin, BISON, and Sockeye applications to resolve the various physics that are essential for the safety analysis of this type of reactor system. The core specifications includes tristructural isotropic fuel, graphite monolith, graphite reflectors, and drums composed of graphite and B 4 C. No moderator or burnable poison pins are used in the design. The fuel enrichment is reduced to control excess reactivity in the core. This core design is not optimized and only serves for testing purposes, since the primary objective of this work is to exercise the multiphysics coupling for this type of reactor system. A three dimensional (3D) core heterogeneous Griffin discrete ordinates (SN) transport model allows the precise calculation of the flux distribution and pin powers. Griffin transfers the power density distribution and obtains a temperature distribution to and from BISON. The BISON model com- putes the 3D core temperature distribution and is coupled to 876 Sockeye subapplications running a heat pipe model. This 3D conduction model is coupled to the various heat pipes via heat flux boundary conditions. The model includes a small gap between the heat pipe and the monolith. Convective heat transfer boundaries with either ambient temperature or condenser temperature as heat sinks are imposed at the model boundaries. The 2D Sockeye heat pipe model uses a vapor- only methodology, which provides the needed resolution for transient calculations and allows the determination of various heat pipe limits. This approach is superior to the superconductor model traditionally used in steady-state calculations. BlueCRAB computes steady-state power and temperature distributions that serve as the initial condition for a loss-of-heat-sink transient simulation. The steady-state results show significant peaking due to the position of the control drum, but this is a characteristic of the particular design used, which is not optimized at this stage. The transient results show the reactor power slowly stabilizing towards a 3% power level after the partial loss of secondary heat removal. Several recriticalities are observed due to cooling through the secondary system but the reactor is self-stabilizing and behaves as expected.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗