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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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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

New ASME Section III, Division 5 Creep-Fatigue Design Rules Based on EPP and SMT Approaches

The integrated elastic-perfectly plastic (EPP) and simplified model test (SMT) creep-fatigue (CF) design methodology, referred to as the EPP-SMT method, is being developed as an alternative for CF evaluation in the design of pressure boundary components for high-temperature reactors. This report reviews the conceptual basis of the EPP-SMT methodology, summarizes the SMT experimental development efforts and results, and provides the technical basis for finalizing the EPP-SMT CF design curves for Alloy 617, based on a combined experimental and numerical approach conducted in FY 2024. This report presents the effect of hold time on the CF design curves for Alloy 617 at elevated temperatures. It includes proposed EPP-SMT CF design curves and tabulated values for continuous cycling, along with the effects of maximum hold time, for the use of this EPP-SMT CF evaluation method.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Advanced Manufacturing Techniques and Compositions of High Entropy Alloys for Nuclear Applications

In line with the objectives of the Department of Energy, Office of Nuclear Energy, Advanced Materials and Manufacturing Technologies (AMMT) program, this work focuses on new materials development and qualification research and development for next-generation, high-temperature nuclear reactors. High entropy alloys (HEAs) have the potential to serve in these extreme environments of next-generation nuclear reactors because of their unique phase transformation pathways and nanoscale and mesoscale microstructures. The current work focuses on understanding such nuclear-energy-relevant HEAs through a detailed literature survey, selected experimental work, and developing a decision matrix with criteria for the identification of HEAs that may have the most impact and value for further examination.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Geochemical and Microbial Dynamics of Hydrogen in a Methane Storage Reservoir

Hydrogen has been identified as a flexible energy carrier with zero or negative emission across multiple energy systems, and existing natural gas infrastructure could be leveraged if hydrogen gas (H2) was blended with methane (CH4). For example, subsurface methane storage reservoirs could be slightly modified to also store hydrogen if a methane/hydrogen blend were injected. However, the compatibility of methane storage reservoirs to include H2 injection has not been fully demonstrated, and this could lead to geochemical and microbiological reactions that alter the reservoir and stored gas content. It is essential that we understand the impact of H2 gas on the biogeochemistry of subsurface storage reservoirs before deploying large-scale H2-CH4 storage, We collected produced fluid from two separate methane storage reservoirs in the Southwestern US. First, we completed a baseline analysis of the biogeochemistry through qPCR, 16S rRNA sequencing, metagenomic sequencing, and geochemical analysis. Each reservoir was found to have unique geochemical conditions and a unique microbial community structure, with Site 1 having a higher TDS and an abundance of Shewanella and Site 2 having a lower TDS and high abundance of Eubacterium and Acetobacterium. Next, we ran a series of high pressure, high temperature reactors under hydrogen storage conditions with the biological sample from one of the storage reservoirs and a 20% H2-80% CH4 gas blend for up to 7 days. Our results show a decrease of hydrogen by 5% in reactors as early as 1-3 days. Previous hydrogen storage work has linked subsurface microorganisms with methanogenesis hydrogen sulfide production, acid production, and microbial corrosion. Our results show minimal change in the fluid chemistry, with the exception of a decrease in dissolved sulfate concentrations. Taxonomic sequencing demonstrated the presence of microorganisms capable of iron redox, acid generation, and hydrogen sulfide production throughout the reactors, suggesting microbial hydrogen consumption may occur through various metabolic pathways. This work demonstrates that site-specific geochemistry and microbiology may impact the efficiency of hydrogen storage in methane storage reservoirs.

environmental microbiology↗
DOE OSTI2025

FY25 Status Update on A709 Code Case Testing at ANL

This report provides an update on the status of the creep rupture testing on the precipitation-treated (PT) Alloy 709 samples fabricated from the first, the second and the third commercial heats, in support of the first, 100,000-hour Alloy 709 Code Case development under the American Society of Mechanical Engineers (ASME) Section III, Division 5 for high-temperature reactor construction. In Fiscal Year (FY) 25, no new tests were initiated, while 5 previously loaded tests were completed. This report presents an overview of the creep-rupture data gathered at Argonne National Laboratory (ANL) since FY21 that totaled 52 completed tests. Metallographic images showing crack morphologies in some representative samples ruptured in FY25 are also included. Key conclusions include: 1) the creep behavior of the three heats of Alloy 709 can be well described by the Larson-Miller relationship; 2) higher temperatures and larger stresses lead to shorter creep life and higher minimum creep rate; 3) Alloy 709 is overall very ductile under creep; 4) the crack morphologies near fracture surfaces are dependent on material’s grain size and test conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Complete Initial Studies in Microstructural Studies of Nuclear Graphite

Nuclear graphite is a critical material for the performance and safety of high-temperature reactors, making its study essential for the successful deployment of next-generation reactors in the United States. During the past year, Oak Ridge National Laboratory (ORNL) has contributed extensively to advancing the understanding of nuclear graphite’s properties, behavior, and applications; particular focus has been given to irradiation effects, oxidation mechanisms, thermal and mechanical performance, and long-term stability under reactor conditions. This report provides an update on the ongoing publications being prepared at ORNL. These studies are collectively intended to build a comprehensive knowledge base on nuclear graphite. The studies focus on key aspects such as microstructural characterization, oxidation behavior, advanced analytical techniques, molten salt infiltration, technique development, and billet-to-billet variability. By summarizing the progress of these publications, this report outlines ORNL’s contributions to the field, identifies knowledge gaps, and proposes future research directions. The findings presented in these studies will help enhance the understanding of nuclear graphite’s behavior, support the qualification of nuclear graphite materials, further the development of new techniques and guidelines for using microscopy techniques, and improve reactor safety and performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Experimental Results on SS 304H and 2.25Cr-1Mo in Support of Developing Inelastic Constitutive Models

To address critical data gaps necessary for updating viscoplastic constitutive material models for Class A materials used in inelastic design analysis under the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors, Oak Ridge National Laboratory (ORNL) conducted experimental studies on stainless steel 304H (SS 304H) and 2¼Cr-1Mo steel (F22), to probe their temperature dependent deformation behaviors. These studies covered a range of temperatures up to the maximum limits specified in Division 5 and are intended to support the development of material models by collaborators at Argonne National Laboratory (ANL).

36 MATERIALS SCIENCE↗
DOE OSTI2025

Advanced Reactor Designs Security Analysis, Risk, and Recommendations: Risks, Consequences, and Possible by-Design Mitigation Approaches Associated with Select Advanced Reactors

Next-generation advanced reactors (ARs) incorporate enhanced safety systems, have smaller source terms, and feature compact modular designs, which should lessen their collective risk profiles. However, to fully evaluate risk, security needs to be a part of the equation. Without taking security into consideration, safety systems and components in the new ARs may be vulnerable to sabotage. These base attributes, coupled with enhanced security features specific to AR design through sound engineering and security-by-design (SeBD), should provide developers and operators with lower inherent security risk profiles. Building security early into the AR design may remove or passively secure potential critical targets from an adversary’s reach , thereby increasing overall safety and security. An integrated approach and diverse design team that includes engineering, operations, and security experts are fundamental to building security into the design without sacrificing fundamental operational efficiencies and principles. The objective of this project was to evaluate the security and safety interfaces for five classes of reactors, identify potential security vulnerabilities of structures, systems, and components (SSC), and underscore the need to consider security alongside safety in the design o f these concepts. The five reactor classes evaluated in this project and presented in this report are molten-salt reactors (MSR), high temperature gas reactors (HTGR), sodium-fast reactors (SFR), advanced light-water reactors (ALWR), and microreactors. These designs were selected because they reflect the concepts that are closest to market deployment and have received significant resource investments from the public and private sector. This project assesses the inherent security risks posed by common classes of ARs, provides a methodology and framework to assess security along with safety, and offers an analysis of potential mitigation strategies that could be incorporated. For each AR technology, the SSCs that relate to radionuclide source safety functions are discussed to understand the SSC contribution to safety and relative importance in the protective strategy for the design. The assumptions that went into evaluating each reactor concept originated from generic publicly available nonproprietary information and should not directly be used to qualify an absolute risk profile nor to rank specific AR designs. Instead, the purpose of the analysis is to understand and compare the generic inherent security risks of different AR technologies.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2024

Fatigue and Creep-Fatigue Evaluation of Alloy 709 at 760 and 816°C

A significant research and development effort is underway to support the qualification of Alloy 709 as a Class A construction material in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors. This initiative includes a comprehensive Alloy 709 code qualification plan aimed at generating extensive material testing data crucial for compiling the code case data package. The data package is essential in establishing material-specific design parameters for Alloy 709 to be used as Section III, Division 5 Class A construction material for fast reactors, molten salt reactors and gas-cooled reactors. An ASME Section III, Division 5 material code case requires the evaluation of mechanical properties from a minimum of three commercial heats, covering anticipated compositional ranges. A key part of the data package involves fatigue and creep-fatigue testing at elevated temperatures, needed for developing the fatigue design curves and the damage envelope of the creep-fatigue interaction diagram (D-diagram). This paper summarizes the strain-controlled fatigue testing on three commercial heats of Alloy 709 at 760 and 816°C with strain ranges between 0.25% and 3%. The fatigue failure data are used to generate a preliminary fatigue design curve. Additionally, the creep-fatigue testing results at 816°C with tensile hold times of 10, 30, and 60 minutes are presented in support of developing the D-diagram for Alloy 709.

Wang, Yanli↗
NASA NTRS1976

Thermal energy storage material thermophysical property measurement and heat transfer impact

The thermophysical properties of salts having potential for thermal energy storage to provide peaking energy in conventional electric utility power plants were investigated. The power plants studied were the pressurized water reactor, boiling water reactor, supercritical steam reactor, and high temperature gas reactor. The salts considered were LiNO3, 63LiOH/37 LiCl eutectic, LiOH, and Na2B4O7. The thermal conductivity, specific heat (including latent heat of fusion), and density of each salt were measured for a temperature range of at least + or - 100 K of the measured melting point. Measurements were made with both reagent and commercial grades of each salt.

Tye, R. P.↗
NASA NTRS1982

Bipulsating Technique for Silicon Production

Method controls reaction temperature and rate of reaction of sodium and silicon tetrafluoride by alternately adding measured amounts of reactants. Technique used in large reactor, where heat dissipation becomes serious problem, to control reactor temperatures. Highly efficient method, which would utilize almost 100 percent of raw materials.

Sanjurjo, A.↗
NASA NTRS1992

Surfactant studies for bench-scale operation

A phase 2 study was initiated to investigate surfactant-assisted coal liquefaction, with the objective of quantifying the enhancement in liquid yields and product quality. This publication covers the first quarter of work. The major accomplishments were: the refurbishment of the high-pressure, high-temperature reactor autoclave, the completion of four coal liquefaction runs with Pittsburgh #8 coal, two each with and without sodium lignosulfonate surfactant, and the development of an analysis scheme for the product liquid filtrate and filter cake. Initial results at low reactor temperatures show that the addition of the surfactant produces an improvement in conversion yields and an increase in lighter boiling point fractions for the filtrate.

Hickey, Gregory S.↗
NASA NTRS2007

Cyanobacteria for Human Habitation beyond Earth

In light of the President s Moon/Mars initiative, lunar exploration has once again become a priority for NASA. In order to establish permanent bases on the Moon and proceed with human exploration of Mars, two key problems will be addressed: first, the production of O2 and second, the production of methane (CH4). While O2 is required for life support systems (LSS), both liquid O2 and CH4 are needed as an oxidizer and a propellant, respectively for the Lunar Surface Access Module (LSAM) and the Crew Exploration Vehicle (CEV). Unlike previous propulsion systems, the new CEV will use liquid oxygen (LO2) as an oxidizer and liquid methane (LCH4) as a propellant. Existing technology (e.g. hydrogen reduction) for the production of liquid oxygen from lunar regolith is very energy intensive and requires high temperature reactors. We propose an alternative approach using iron-tolerant cyanobacteria. We have found that iron-tolerant cyanobacteria (IT CB) are capable of etching iron-bearing minerals, which may lead to bonds breaking between Fe and O of common lunar mare basalt Fe-oxides including ilmenite, pseudobrookite, ferropseudobrookite, and armalcolite with the subsequent release of both Fe, Ti and oxygen as byproducts. We also propose to use CB biomass for CH4 production as carbon stock and a propellant. Both processes can be accomplished in an energy and cost effective manner because sunlight will be used as an energy source and allows the reactions at ambient temperatures between 10-60 C. Current evaluations include assessing the thermodynamics of such biogenic reactions using a variety of nutrients and atmospheric parameters, as well as assessing the rates and species variation effects of the driving reactions.

Brown, Igor↗
NASA NTRS2006

Iron-Tolerant Cyanobacteria for Human Habitation beyond Earth

In light of the President's Moon/Mars initiative, lunar exploration has once again become a priority for NASA. In order to establish permanent bases on the Moon and proceed with human exploration of Mars, two key problems will be addressed: first, the production of O2 and second, the production of methane (CH4). While O2 is required for life support systems (LSS), both liquid O2 and CH4 are needed as an oxidizer and a propellant, respectively for the Lunar Surface Access Module (LSAM) and the Crew Exploration Vehicle (CEV). Unlike previous propulsion systems, the new CEV will use liquid oxygen (LO2) as an oxidizer and liquid methane (LCH4) as a propellant. Existing technology (e.g. hydrogen reduction) for the production of liquid oxygen from lunar regolith is very energy intensive and requires high temperature reactors. We propose an alternative approach using iron-tolerant cyanobacteria. We have found that iron-tolerant cyanobacteria (IT CB) are capable of etching iron-bearing minerals, which may lead to bonds breaking between Fe and O of common lunar mare basalt Feoxides including ilmenite, pseudobrookite, ferropseudobrookite, and armalcolite with the subsequent release of both Fe, Ti and oxygen as by-products. We also propose to use CB biomass for CH4 production as carbon stock and a propellant. Both processes can be accomplished in an energy and cost effective manner because sunlight will be used as an energy source and allows the reactions at ambient temperatures between 10-60 C. Current evaluations include assessing the thermodynamics of such biogenic reactions using a variety of nutrients and atmospheric parameters, as well as assessing the rates and species variation effects of the driving reactions.

Brown, Igor↗
NASA NTRS2015

Mars Propellant Production with Ionic Liquids Project

This project seeks to develop a single vessel for carbon dioxide (CO2) capture and electrolysis for in situ Mars propellant production by eliminating several steps of CO2 processing, two cryocoolers, a high temperature reactor, a recycle pump, and a water condenser; thus greatly reducing mass, volume, and power.

Falker, John↗
DOE OSTI2025

Open Architecture for Cost Savings in Advanced Nuclear Reactors

Recently, nuclear power plant build projects in the West have run over budget due to high capital costs and schedule overruns. Compared to other sources of energy, nuclear power plants have higher capital costs. Reactors are often different at every site, resulting in a lack of standardization. Nuclear is expected to compete with other low carbon sources of energy which have lower capital costs making it essential for nuclear to develop ways of reducing costs. Strategies such as standardization, learning rates, modularization, and schedule reduction in advanced reactors can reduce nuclear costs by about 40%. Standardization as a way of cutting capital costs has been explored even in large nuclear power plants. Standardization of certain plant components can result in lower component and installation costs and higher learning from experience. Standardization can be achieved by adopting a criterion of key performance indicators and general design principles for a specific system or component such as the balance of plant. Modularization allows the construction of certain components of SMRs in a factory, which saves time, increases productivity, and encourages higher learning rates. Production learning decreases the time and the cost related to an activity. The potential for modularized components of advanced reactors to be manufactured in factories makes it conducive to achieving higher learning rates. Developing large-capacity nuclear programs through sequential builds cultivates a higher learning rate, which in effect may reduce schedule overruns. Open architecture has been identified as a way to drive standardization among advanced reactor designs and result in cost savings. Open architecture (OA) is defined as a design enabling a diverse supply chain by defining and publishing requirements of systems or equipment in functional and/or interface terms, utilizing technical standards in widespread use. Currently, the nuclear industry’s approach is to use closed architecture, making most designs proprietary. However, collaboration between various advanced reactor vendors and suppliers utilizing the concept of open architecture can result in modular and standardized architecture of subsystems or subcomponents of a nuclear power plant. Completely standardizing nuclear power plants may be impossible, however, certain common subsystems amongst the various reactor designs could be standardized and/or access a wider supply chain and leverage existing learning from other sectors. Open architecture will save time and allocate resources to the parts of the plants that have the most unique features. A key advantage of open architecture is its ability to improve production learning across advanced reactors (AR) types in the industry, by providing and utilizing the same kind of component. Sodium fast reactor (SFR), High Temperature Gas Reactor (HTGR) and Molten Salt Reactor (MSR) are the advanced reactors considered for this project. This paper aims to determine the cost savings in advanced reactor programs due to open architecture learning rate. This work is an extension of work done on light water reactor small modular reactors; the cost methodology was utilized to investigate the impact of open architecture on advanced reactors with a particular focus on sodium fast reactors. The cost data on sodium fast reactors used in the model presented the most adequate information required for the analysis.

Advanced Nuclear Reactors↗
DOE OSTI2024

Evaluation of the Simplified Assessment Peak Equivalent Stress Design Limit Probability of Failure

The ASME Boiler and Pressure Vessel Code Section III Division 5 Subsection HH Subpart A (HHA) outlines requirements and qualification methods for graphite core components in high temperature reactor environments. One such structural qualification method is the Simplified Assessment outlined in HHA-3220, which represents a comparison of service level specific equivalent stresses to allowable stress values derived from target probability of failures (POFs). The target POF is based on the structural reliability class (SRC) for the components and service condition of interest, and is used to establish design limits for combined membrane and peak equivalent stresses. Combined membrane stress limits are derived from two parameter Weibull distribution fits of the graphite grade specific tensile strength data and the target POF. The combined membrane stress limits are then scaled based on the ratio of flexural strength to tensile strength (referred hereinafter as Rtf) to establish the corresponding peak equivalent stress limits. It is desired that the target POF is at least maintained for both combined membrane and peak equivalent design limits, especially with the use of Rtf. The assessment presented herein examines tensile and flexural data for medium and fine grade graphites relative to the calculated combined membrane and peak equivalent stress limits per HHA-3220 and illustrates the simplified assessment method results in anticipated, or at least conservative, peak equivalent design stress limits.

36 - MATERIALS SCIENCE↗