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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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Kinetics and Reactor Modeling of VFA Ketonization for Sustainable Aviation Fuel Production

Ketonization of volatile fatty acids (VFAs) produced via arrested methanogenesis of wet waste represents the only unit operation of wet waste upgrading to sustainable (net-zero or negative life cycle CO2 emissions) aviation fuel not currently in industrial practice. Ketone product yields of close to 100% have been obtained during gas-phase reactions of VFAs over oxide catalysts at the laboratory scale, but design of an industrial ketonization reactor requires understanding of the impacts of reactant and product partial pressures, deactivation, and heat and mass transport phenomena on observed ketonization rates. Our work leverages rigorous kinetic analysis of ketonization processes to inform reactor scale-up efforts through packed-bed reactor modeling. We first present results of a kinetic study of ketonization of a model VFA, hexanoic acid, over an industrial ZrO2 catalyst performed in a packed-bed microreactor in conditions free of significant heat or mass transfer gradients. Major findings of the analysis include: (i) hexanoic acid saturates all catalyst active sites at relatively low partial pressure (~10 kPa) and (ii) ketonization products 6-undecanone, water, and CO2 inhibit reaction rates. Kinetic data are used to fit a rate expression quantifying the functional dependence of ketonization rate on partial pressures of VFA reactants and ketone, water, and CO2 products. A packed-bed reactor model describing vapor-phase hexanoic acid ketonization over ZrO2 extrudate catalysts is developed based on the kinetic model. The effects of (i) bed- and pellet-scale mass- and heat-transfer limitations and (ii) axial pressure drop guide our development of recommendations for optimal sizing, temperature, and influent flow composition of an industrial-scale ketonization reactor. The quantitative understanding of VFA ketonization developed in this study represents an advance toward derisking the VFA ketonization step of wet waste upgrading to sustainable aviation fuel.

biojet↗
DOE OSTI2021

Assessment of Nuclear Data Needs for Advanced Reactor Demonstrations: Application to the Molten Chloride Reactor Experiment (MCRE)

A number of advanced reactor concepts are planned for near-term demonstrations including microreactors, larger demonstrations and space nuclear systems. These reactor concepts are based on a wide variety of reactor technologies, including sodium, gas, and salt cooling. An overlooked area in the development and ultimate startup of these reactors is addressing nuclear data needs that allow confident prediction of the criticality, safety requirements, and operation of the reactors. In this paper we try to address the problem of assessing nuclear data needs and possible remedies to reduce the existing uncertainties for advanced nuclear reactors. A methodology for defining these needs is described. The case of the Molten Chloride Reactor Experiment (MCRE) has been considered and the related investigation highlights the specific needs for reducing uncertainty on the 235U capture and 35Cl (n,p) reactions. Integral experiments relatively inexpensive are indicated as possible solution for significantly reduce the current associated uncertainties.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

U.S. High Temperature Materials Highlights

U.S. GIF VHTR work is continuing on graphite qualification, Alloy 617 regulatory issues beyond the Code space, Alloy 800H weldments, and ASME Codes and Standards R&D is still considering both pebble bed and prismatic and steam generator and heat exchanger U.S. DOE Advanced Reactor Demonstration Program (ARDP) Two U.S.-based teams were selected to demonstrate advanced nuclear reactors in the United States that can be operational by 2027 One of the teams is X-energy (Rockville, MD) which will demonstrate a modular gas-cooled reactor design (Xe-100) with four 80 MWe, TRISO fuel, pebble bed reactors A number of U.S.-based teams were selected to design and develop safe and affordable reactor technologies that can be licensed and deployed over the next 10 to 14 years (Risk Reduction) One of the teams is BWXT Advanced Technologies, LLC which will develop a commercially viable transportable microreactor with the design focused on using TRISO fuel particles and silicon carbide (SiC) matrix A number of U.S.-based teams were selected to assist the progression of advanced reactor designs in their earliest phases (Advanced Reactor Concepts-20) One of the teams is Massachusetts Institute of Technology which will mature the Modular Integrated Gas-Cooled High Temperature Reactor (MIGHTR) concept with a horizontal compact design from a pre-conceptual stage to a conceptual stage to support commercialization

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Demonstration of ACCERT Software for Nuclear Power Plant Techno-Economics

In the past few years, there has been a renewed interest in the deployment of nuclear power for decarbonizing the electricity grid as well as a range of industrial applications. As the demonstrations of advanced nuclear power plants start to begin, there will likely be a further increase in this interest. As nuclear is being considered as a part of the energy mix, understanding the cost of nuclear energy becomes increasingly important for all stakeholders including advanced reactor vendors (for making design decisions and marketing their designs), users of nuclear energy (e.g., to estimate the cost of decarbonization of other industries using nuclear), and government (e.g., in capacity expansion models that are used in framing policy). In this summary, we demonstrate a software tool called ACCERT that is currently being developed with funding from the Systems Analysis and Integration (SA&I) program under the Department of Energy’s Office of Nuclear Energy (DOE NE). ACCERT is a cost estimation and techno-economics tool for nuclear power plant applications that includes a database of (a) cost estimates of various ‘reference’ nuclear power plant designs gathered from existing literature, and (b) algorithms developed from these costs that can be used extrapolate the existing costs and perform a bottom-up cost estimation of other designs. A companion summary describes the software and its design in more detail and this summary presents a demonstration for four different nuclear power plant designs: a pressurized water reactor (PWR), high-temperature gas reactor (HTGR), sodium fast reactor (SFR), and a heat-pipe microreactor. The demonstrations include the reference cost estimates and the cost estimates of a modified design for each reference case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The Deimos Experiment: Advanced Reactor Testbed

Advanced reactor initiatives are growing significantly through programs nationwide. This research area includes small modular reactors, microreactors, and space reactors. Many of the reactors being designed are untested concepts. They include unique moderators, varying fuel types, high temperatures, and compact configurations. The shift in fuel type, from highly enriched uranium (HEU) to high assay, low enriched uranium (HALEU), is particularly important as it has driven many of the other changes. For example, lower enrichment requires advanced moderators, which in turn require different reflectors to make the systems compact. The change in materials including the transition from HEU to HALEU affects the temperature feedback of the systems. Additionally, these advanced reactor concepts generally have a thermal neutron spectrum in contrast to earlier fast spectrum advanced reactor. With the extensive changes from previous reactor designs, validation experiments are needed. The National Criticality Experiments Research Center (NCERC) is uniquely equipped to perform such experiments. The Deimos experiment, designed for execution at NCERC, will serve as a testbed for advanced reactor concepts. It will use HALEU fuel in a graphite matrix, provide the ability to use advanced moderators, and allow measurements of temperature reactivity coefficients (TRCs).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

OECD-NEA HTTF Benchmark Progress and Updates

General slides discussing High Temperature Test Facility benchmark progress, Prismatic HTGR deployment, modeling and simulation tools for validating systems, and verification and validation issues. Also discuss collaboration and conclusions of RELAP5-3D validation activities based on HTTF with trends in data and international impact to accelerate deployment of prismatic HTGR microreactors by providing an opportunity for designers to assess their codes against experimental data and solutions from other codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

National Reactor Innovation Center Annual Report FY 2023

It is my privilege to present this year’s annual report for the National Reactor Innovation Center (NRIC). Now rounding out our fourth year of operation, NRIC is poised for successfully delivering our most foundational promise: to accelerate the testing and demonstration of advanced nuclear technology by providing access to national laboratory assets and expertise. Some specific goals include: • Establish and maintain four new experimental facility capabilities: o MSTEC – 2025 o METL – Operational o HeCTF – Operational o Creep Frames – 2026 • Construct and operationalize two advanced reactor test beds (DOME and LOTUS) for integrated technology demonstrations and experimentation – 2028 • Complete minimum of two advanced nuclear technology tests – 2030 o Advanced Construction Technology – 2025 o Advanced Microreactor test in DOME or LOTUS – 2027 & 2029

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

MARVEL Project Code of Record, COR-0011

MARVEL Project Code of Record identifies the codes, standards, and procedures necessary to design, develop, construct, and startup the Microreactor Applications Research Validation and Evaluation (MARVEL) Project. No code exceptions to construction of the MARVEL reactor in the traditional sense were identified, meaning there are no technical aspects that fail to meet a code requirement. However, due to the novel design of the MARVEL reactor, not every aspect or system can be precisely covered by the existing ASME Boiler and Pressure Vessel Code. In those cases, the document identifies (with concurrences received) how the ASME Boiler and Pressure Vessel Code is applied in an equivalent or analogous manner.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

An Initial Evaluation of the Elevated-Temperature Cyclic Properties of Optimized 316H Stainless Steel Fabricated by Powder Metallurgy Hot Isostatic Pressing

Powder metallurgy hot isostatic pressing (PM-HIP) is a mature advanced manufacturing technology, which can consolidate metallic powder under high temperature and pressure to generate near-net shape components. This process can lower production costs and reduce component lead times for nuclear reactor construction and can be readily deployed in the near term for microreactor components because of their smaller size. The purpose of this work is to continue the evaluation of elevated-temperature cyclic material properties for PM-HIP 316H stainless steel. Prior scoping results have shown that the commercially procured PM-HIP stainless-steel materials have reduced creep-fatigue performance compared to wrought 316 stainless steels. To better understand how PM-HIP processed material properties compare to traditional manufacturing methods, additional billets of PM-HIP 316H stainless steel were procured and tested. This testing is necessary to understand how elevated-temperature properties are influenced by the PM-HIP process and what data are needed for incorporating PM HIP 316H into the American Society of Mechanical Engineers Boiler Pressure Vessel Code, Section III, Division 5 for high temperature reactor construction. Specifically, an additional PM-HIP 316H stainless steel billet with lower oxygen and nitrogen concentrations was fabricated by the Nuclear Advanced Manufacturing Research Centre of the United Kingdom and tested to understand if elevated-temperature cyclic properties can be improved. The creep-fatigue resistance of the material from this new billet did not show appreciable improvement compared to prior testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermal Integration of DETAIL Systems

A thermal connection between the Microreactor AGile Non-nuclear Experimental Testbed (MAGNET) and the thermal energy distribution system (TEDS) has been a key part of the Dynamic Energy Transport and Integration Laboratory (DETAIL) plan since the development of the Integrated Energy Systems Roadmap in 2020. This connection provides an expanding demonstration capability for integrated system operations. The thermal connection between MAGNET and TEDS required the selection of a heat exchanger and valves to isolate flow. Research staff within the NEET Crosscutting Technology Development program selected equipment after receiving proposals from multiple vendors. A helical coil heat exchanger from Graham Corporation was selected for its small volume relative to its heat transfer surface area to fit in the limited space available. Triple offset butterfly valves with pneumatic actuators from Flowserve were selected to isolate and/or control flow of the hot gas through the heat exchanger. Staff from Idaho National Laboratory’s (INL) Facilities and Site Services Engineering department designed the piping and structural support. A mechanical construction firm was contracted to install the system. This report documents the design and construction of this thermal connection.

25 ENERGY STORAGE↗
DOE OSTI2023

National Reactor Innovation Center Update

Accelerate the testing and demonstration of advanced nuclear technology by providing access to national laboratory assets and expertise: Establish & maintain four new experimental facility capabilities MSTEC – 2025; METL – Operational; HeCTF – Operational; Creep Frames – 2026 Construct and operationalize two large reactor test beds for integrated technology demonstrations and experimentation - 2028, Complete minimum of two advanced nuclear technology tests – 2030 Advanced Construction Technology – 2025 Advanced Microreactor test in DOME – 2027 & 2029

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

ECAR-5139 MARVEL Radiological Source Term

As part of the Microreactor Applications Research Validation and Evaluation Project (MARVEL) design effort, a preliminary end of life reactor fuel radiological source term is calculated in support of the MARVEL preliminary hazard evaluation. The source term calculation assumes a two-year uninterrupted operating life at a power of about 111 kW for a total energy of 7,000,000 MJ. A final source term for the final hazard evaluation will need to be generated or an evaluation performed to determine the applicability of the source term documented here for use in the final hazard evaluation.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

RELAP5-3D Solutions to Exercise 1 of the OECD-NEA HTTF Benchmark

The high-temperature gas-cooled reactor (HTGR) is an advanced reactor concept that has received considerable attention over the last 60 years. HTGRs boast high coolant outlet temperatures, passive safety, and large margins to fuel failure. Recent years have seen increasing interest in HTGRs, whether large-scale HTGRs for electricity or process heat applications or as microreactors serving remote communities. HTGRs have considerable potential to fulfill these applications, but modeling and simulation tools typically used for reactor safety analysis have not been validated for HTGR modeling. The High-Temperature Test Facility (HTTF) was constructed at Oregon State University (OSU) to provide validation data for HTGR modeling. Recently, Idaho National Laboratory (INL), OSU, Argonne National Laboratory, Canadian Nuclear Laboratories, and the University of Tennessee Knoxville have collaborated to develop an HTGR thermal hydraulics benchmark based on HTTF experiments. This summary presents results from a portion of that benchmark using RELAP5-3D

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Physics demonstration and verification of MOOSE framework reactor module meshing capabilities

The recently developed Reactor module in the open-source MOOSE framework includes finite element meshing capabilities for analysis of common reactor geometries. Capabilities in the Reactor module have been employed to generate meshes for physics applications including a sodium-cooled fast reactor core analysis using Griffin, a fast reactor assembly thermal deformation analysis using MOOSE Tensor Mechanics, and a heat-pipe cooled microreactor coupled analysis using Griffin, Bison, and Sockeye. The process to build these meshes using MOOSE's meshing capabilities is described. Physics simulation results using MOOSE-based meshes have been verified to match results which leverage external meshing software such as Cubit or Argonne's Mesh Tools. MOOSE's Reactor module provides significant advantages compared to the use of external meshing tools when analyzing Cartesian and hexagonal reactor lattices using MOOSE-based applications: accessibility to the end user, low barrier to entry for new users, speed of mesh generation, volume preservation of meshed fuel pins, and simplification of analysis workflow when using MOOSE applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Recent Advanced Reactor Multiphysics Model Highlights in the Virtual Test Bed (VTB)

The Virtual Test Bed (VTB) host over 30 distinct simulations that showcase state-of-the art capabilities across the national lab complex. An update on the status of models on the VTB is summarized here, along with a more detailed overview of select recent new capabilities to showcase. All of the major advanced reactor types are represented in the VTB. The first example consists of a multiphysics simulation to track the transport of species in Molten Salt Reactors using depletion, advection, and thermochemical calculations. The second consists of a coupled neutronic and thermal hydraulic simulation to validate a gas cooled reactor. The third consist of pebble-bed equilibrium model for a fluoride high-temperature reactor. The fourth is a high-fidelity neutronic and thermal hydraulic model of a liquid metal reactor assembly. And lastly the fifth consists of transient multiphysics simulations of heat pipe microreactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Overview of Virtual Test Bed (VTB)

The Virtual Test Bed (VTB) was established to support the deployment of advanced reactors by facilitating the use of state-of-the-art modeling and simulation capabilities. These tools can solve previously prohibitive challenge problems with greater levels of fidelity and in a multiphysics framework. A VTB repository was recently set up to host models for a wide variety of reactor types and using a range of Nuclear Energy Advanced Modeling and Simulation-developed software. The repository currently includes models for High-Temperature Gas-Cooled Reactors, Sodium Fast Reactors, Fluoride-salt-cooled High-temperature Reactors, Molten Salt Reactors, and Heat-Pipe Microreactors. This article provides an overview of the capabilities currently in the VTB repository.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Development of Shielding Benchmarks Using the Godiva IV Assembly

The Nuclear Criticality Safety Program (NCSP) is developing a shielding benchmark using the Godiva IV assembly as a source. Its present status is reviewed herein. Even eight decades into the nuclear era, substantial work remains to develop a database of shielding benchmarks to support future nuclear development. A nuclear simulation is only as good as its supporting data, inputs, and validation basis. Uncertainty in these areas is addressed using conservatism, which adds margin and, occasionally, cost. In many shielding situations, high accuracy is not necessary because additional material is not particularly expensive. After all, 1–2 cm of lead often reduces the gamma ray dose substantially. However, in certain areas, conservatism can add unnecessary cost. These areas include mobile shielding applications such as casks, ships, microreactors, and spacecraft, where weight and, thus, margin is expensive. Although these characteristics are side benefits for NCSP shielding benchmark development, the main driver is enabling more reliable placement of criticality accident alarm systems (CAASs) in nuclear material facilities, such as those dedicated to the production of advanced reactor fuels. CAAS placement relies on more than accurate data and code validation. It also relies on sufficiently accurate materials specifications, geometry specifications, and a well-defined, alarm-producing baseline accident. All these things require tacit knowledge and understanding of the problem being evaluated. Benchmarks can help ensure this understanding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗