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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

FENIX: An Open-Source Multiphysics Integrated Framework Enabling Collaborative Development of Plasma Facing Component Modeling Capabilities

Advanced modeling and simulation tools have a crucial role to play in accelerating fusion energy deployment as a sustainable power source. Multiphysics, high-fidelity computational tools can help understand, model, and quantify the complex interactions between materials performance, plasma and neutron exposure, and engineering processes. As a result, they accelerate the design, safety analysis, and performance evaluation of fusion systems. This webinar introduces the Fusion ENergy Integrated multiphys-X (FENIX) framework, an open-source multiphysics tool for plasma facing component modeling. FENIX leverages the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, which has been developed by the United States Department of Energy Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. FENIX couples various MOOSE capabilities such as heat transfer, thermomechanics, thermal hydraulics, electromagnetics, and plasma kinetics with the MOOSE-based applications Cardinal (neutronics) and TMAP8 (tritium transport). During the webinar, we will present FENIX and discuss how its modularity, openness, software quality assurance processes, and licensing approach supports effective collaborations, including public-private partnerships.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Effect of Impurity Atoms on Thermonuclear Yield

The presence of impurity atoms in an inertial confinement fusion (ICF) implosion can greatly affect the burning of the thermonuclear fuel. This is a serious concern for attaining ignition as high atomic number atoms, such as carbon, can mix into the hot spot and reach the deuterium–tritium (DT) fuel, absorb energy from the implosion, and radiate away that energy. The temperature of the DT fuel drops and reduces the reactivity of the fuel resulting in lower thermonuclear yield. Fuel impurities can also be a useful diagnostic as when higher-Z atoms, such as Ar, Kr, and Xe, are purposefully mixed into the fuel. From the spectra emitted by these atoms, the temperature and density of the fuel can be deduced. The long history of using such tracer atoms by the national ICF program shows the great advances in experimental technique, diagnostic instruments, and atomic-physics modeling over the past 50 years. A particular set of experiments, the “High-Z” campaign, is reviewed in depth as this was the most systematic study of tracer element composition and quantity ever reported. Difficulties in interpreting the results of those experiments due to the limited code capabilities at that time are discussed. A main concern for the High-Z campaign was the then current state of non-Local Thermodynamic Equilibrium (nLTE) atomic physics models. The present state of nLTE modeling and its implementation in Los Alamos’s multi-physics codes is discussed in this report. To further the understanding of the complex interaction of high-Z atoms in the gas, a survey of relevant existing experimental platforms at the National Ignition Facility (NIF) and the Omega Laser Facility are presented along with the current state of pertinent diagnostics at NIF and Omega. This review concludes with a discussion of the trade-off available between temperature and volume of the imploded ICF capsule with the goal of maximizing the total yield.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Design Considerations to Ensure Robustness of the ITER Diagnostics Residual Gas Analyzer

Increasing robustness of the ITER diagnostic residual gas analyzer (DRGA) is critical for the potential control of plasma heating and fuel-cycle processing. Robustness is a requirement for a diagnostic to have a control function. The DRGA is a multisensor diagnostic system capable of resolving isotopic compositions of hydrogen and helium as well as other heavier elements and compounds. The divertor-specific DRGA system is intended to measure the composition of gases in the ITER subdivertor region and midplane. Its analysis station will be located in a port cell at the divertor level. From there, it will sample a slip stream of gas from the cryogenic pump duct. It will then exhaust into a shared roughing line where helium or other light gas impurities, some potentially from other diagnostic systems, are likely to be present. In order to provide reliable measurements, the DRGA must be robust in areas such as plasma optical emission source geometry for a compact design, mitigation of back-streaming from light gases and resilience to the ITER port cell environment, and radiation hardening of the DRGA electronics. By incorporating robustness into the design, areas of plasma heating and fuel-cycle control may be explored with the DRGA for ITER and next-generation fusion devices.

Quinlan, Brendan↗
DOE OSTI2020

Laser-Based Means for Accelerating Nuclear Decay Rate

This project aims to replicate and build upon a reported series of results whereby nuclear decay of unstable species has been accelerated under laser irradiation in the presence of resonantly excited plasmonic nanoparticles. We are currently utilizing a laser system similar to that used within these reports as well as optimizing nanoparticle solution environments in an attempt to replicate reported results or provide alternative pathways that may explain some perceived discrepancies. The current work is being performed with equipment and techniques currently available at SRNL, including nanoparticle synthesis and pulsed lasers for plasmon excitation. The results of these experiments could provide quantitative data that can be used to resolve the discrepancies in the calculations and better estimate the ultimate utility of the process. If verified, this would represent a pathway for using readily available and scalable technology to reduce the long-term radiological storage requirements of certain forms of nuclear waste.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Enabling Robust Compressor Operation Under Various sCO 2 Conditions at Compressor Inlet

The supercritical carbon dioxide (sCO 2 ) power cycle is one of the most investigated power conversion systems in last decade. The reason for this is its wide range of potential applications from fossil fuels (Natural gas), nuclear (Fusion, Fission), through concentrated solar power, energy storage system and waste heat recovery systems to power-to-x systems. The main advantage of the sCO 2 power cycle is very similar operation conditions, especially for the compressor, for all potential applications. Because of this the potential cost of the system can be reduced. However, the sCO 2 power cycle has still several issue that needs to address before commercial application. The current key issue is design of the primary compressor. The primary compressor is one of the key components in the sCO 2 power cycle; it operates with inlet conditions just above the critical point of CO 2 to maintain high fluid density and reduce power requirements.

14 SOLAR ENERGY↗
DOE OSTI2024

Kinetic Plasma Simulation in the MOOSE Framework: Verification of Electrostatic Particle In Cell Capabilities

In magnetic confinement nuclear fusion reactors, the interaction between the plasma edge and plasma facing components is extremely important. At the plasma edge, a kinetic representation such as particle-in-cell (rather than a fluid representation) is required to accurately capture the plasma behavior. General purpose particle-in-cell plasma simulation capabilities have been developed in the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. This new capability is a part of the development of a new MOOSE-based framework for modeling plasma facing components, the Fusion ENergy Integrated multiphys-X (FENIX) framework. In this work, the verification of foundational particle-in-cell capabilities in FENIX is presented. This new plasma simulation capability has three main components: moving particles in discrete steps on the finite element mesh, mapping charge density from the particle's location to the finite element mesh, and solving for the electrostatic potential based on the charge density mapped from particles to the mesh. In this paper, simple verification problems demonstrating each of these new capabilities are presented, and future work includes electromagnetic capabilities and Monte Carlo collisions with neutral gas particles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Approach to startup inventory for viable commercial fusion power plant

With the increasing efforts to commercialize fusion power, private and government organizations are investing heavily in the development of technology to support a viable fusion power plant. Deuterium-Tritium (DT) fueled reactors are more prevalent than other proposed designs, requiring tritium processing and handling technology for safe operations and self-sufficiency. Further, each fusion power plant will need a specific-to-design startup inventory of tritium to begin operations. This startup inventory is required prior to breeding and is the minimum tritium inventory required to fill each processing component in the fuel cycle, to offset radioactive decay losses, and to avoid a zero-fuel situation for continuous operation. We present an approach to calculate the startup tritium inventory for a 500 MW th reactor, with considerations for reserve inventory for maintenance and commissioning. A baseline startup inventory was calculated to be approximately 327 gs. This value was obtained using modest assumptions about the technology and operating parameters of a fusion power plant. The required operating reserve inventory or the inventory necessary to keep a fusion power plant operational using only direct internal recycling for 24 h for the same plant design is approximately 642 gs. The approach and findings of this paper will enable fusion energy stakeholders to better utilize the existing scarce global tritium supply.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Applications of fuzzy logic and best-worst method for tritium sensor selection

Accurate assessment of tritium as a fuel source is critical in fusion reactions, necessitating effective sensor evaluation methods. This study investigates a multi-criteria decision-making framework for selecting tritium sensors, integrating fuzzy logic to enhance decision quality. Initial attempts at applying fuzzy logic were found to be too elementary and failed to capture the complexity of multi-criteria selection; this prompted a refined approach that incorporated expert insights and advanced ranking techniques for sensor evaluation. The research used a two-stage methodology. In the first stage, important criteria and sub-criteria for sensor performance were identified and defined. These criteria were then weighted and scored using a fuzzy best-worst method, drawing upon expert opinions to ensure relevance and validity. The second stage involved interpreting information about varying sensors to rank them based on their overall criteria scores, encouraging the selection of the most suitable options. The result of the study is a proposed method for effective sensor selection in fusion reactors, which in turn will significantly improve the reliability of tritium monitoring in fusion applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Comparative Analysis of Confidence Metrics for Nuclear Criticality Safety

Nuclear criticality safety standards provide guidance on the requirements and recommendations to establish confidence in computerized model results used to support operation with fissionable materials. By design, the guidance is not prescriptive, leaving the analysts free to determine how various sources of uncertainties are to be statistically aggregated. This report compares the analyses and key assumptions behind four notable methodologies documented in the nuclear criticality safety literature: the parametric, nonparametric, Whisper, and TSURFER methodologies. Because of the involved use of statistics entangled with heuristic recipes, the results of these methodologies are often difficult to interpret. Also, they are augmented by additional large administrative margins, eliminating the incentive to understand their differences. With the new resurgent wave of advanced nuclear systems focused on economizing operation—including advanced reactors, fuel cycles, and fuel concepts—there is a strong need to develop a clear understanding of uncertainties and their fusion methodologies to reduce uncertainties in a scientifically defensible manner. This report offers a deep dive into the various assumptions of the four noted methodologies, their adequacy, and their limitations, to provide guidance on developing confidence for the emergent nuclear systems. These systems are expected to be challenged by the scarcity of experimental data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Physics Flash Spring 2020

Published by the Physical Sciences Directorate at LANL, the Spring 2020 issue of this newsletter contains an update from the division leader, Tanja Pietraß, Physics Division news roundaup, welcome to new hires, snapshots of Physics managers' work-from-home offices, and service celebrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

MOOSE: A Modular Platform for Fission and Fusion Multiphysics

The Multiphysics Object-Oriented Simulation Environment (MOOSE) Framework, as well as MOOSE-based simulation tools, have accelerated the development of fission energy and advanced reactor technologies through the United States Department of Energy, Office of Nuclear Science, Nuclear Energy Advanced Modeling & Simulation (NEAMS) Program. MOOSE contains a complete platform of multiphysics simulation capabilities, capable of running on massively parallel systems, and is developed in an open-source manner with great attention paid to high-quality software quality assurance practices. This overall approach could greatly benefit the fusion energy community, which requires rapid design iteration and improvement in order to facilitate the successful development of fusion as an alternative energy source to fossil fuels. In the first half of this talk, applications of MOOSE and MOOSE-based tools for advanced reactor designs will be showcased, as well as MOOSE ecosystem infrastructure (such as the NEAMS Virtual Test Bed) that enables and accelerates fission reactor design. In the second half, a discussion of how the MOOSE approach to modeling and simulation is currently being applied internationally in fusion energy research and development at the United Kingdom Atomic Energy Authority will be discussed, and ongoing/future domestic research efforts will be highlighted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

The comparative strength and fracture toughness properties of commercial 95W-3.5Ni1.5Fe and 95W-3.5Ni1.5Cu tungsten heavy alloys

Due to the sharp surface cracks resulting from high and cyclic thermal loads, fracture toughness is a crucial property for fusion divertor materials. Tungsten heavy alloys (WHA) have emerged as a leading candidate for plasma facing component applications. Both the WNiFe and WNiCu WHAs have been considered. However, there is no fracture toughness data in the literature on the WNiCu WHA. Therefore, maximum load elastic–plastic fracture toughness (K Jm ) of a commercial 95W-3.5Ni1.5Cu (NiCu) WHA plate is reported here for the first time. The NiCu WHA K Jm is compared to a corresponding K Jm for commercial 95W-3.5Ni1.5Fe (NiFe) WHA, along with comparisons of their respective microstructures, microhardness, and tensile properties. Although their room temperature yield strengths are similar (≈ 600 MPa), the NiFe WHA has higher ultimate stress (818 MPa vs. 642 MPa) and total elongation (8% vs. 1.2%) compared to the NiCu WHA. Most notably, the room temperature elastic–plastic K Jm for the NiFe WHA of ≈ 80 ± 8 MPa$\sqrt{m}$, is nearly twice that of the NiCu WHA at ≈ 47 ± 4 MPa$\sqrt{m}$. Further, while the NiFe WHA experienced stable ductile tearing, the crack propagated unstably (approximately elastically) after reaching the maximum load in the NiCu WHA. The higher K Jm of the NiFe WHA is attributed to its higher flow stress and ductility, along with the energy dissipation and microcrack dilatational stress shielding, provided by a wide distribution of arrested process-zone microcracks, which are largely absent in the NiCu WHA. Since, the 95W-NiFe WHA has distinctly better mechanical properties than 95W-NiCu WHA, it is a more suitable candidate for the fusion reactor application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

TRITIUM TRANSPORT MODELING AT THE PORE SCALE IN CERAMIC BEEDER MATERIALS USING TMAP8

Fusion reactors depend on the blanket material to breed and release tritium at the same rate or faster than it is consumed by the fusion reaction. Cellular ceramic breeders are dense materials that can maintain a high tritium breeding ratio while promoting tritium release thanks to highly connected pores. Assessing the tritium breeding capabilities of these materials requires a combination of extensive experimental and modeling efforts. In this work, we develop and calibrate a multiphysics model of tritium transport. This novel model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. We perform a sensitivity analysis and calibrate the model by comparing its predictions against experimental measurements of deuterium absorption. The calibrated model is then used to model tritium absorption in samples with different pore microstructures to investigate the effect of pore interconnectivity on tritium absorption. The model is part of the development of the multiscale, multiphysics framework for tritium transport (i.e., the Tritium Migration Analysis Program [TMAP8]), which is itself built on top of the finite element multiphysics framework MOOSE. This study demonstrates some of TMAP8’s capabilities, and is the first step toward assessing the tritium breeding capabilities of ceramic breeder material designs.

08 HYDROGEN↗
DOE OSTI2022

Calibration of a mesoscale tritium transport model for ceramic breeder materials in TMAP8 using experimental data

Due to the scarcity of the long-term external tritium supply, lithium-containing breeder materials are used in the blanket of fusion reactors to produce tritium faster than the deuterium-tritium fusion reaction consumes it. A cellular breeder is a promising material with higher lithium density and thermal conductivity than conventional ceramic pebbles and enhanced tritium extraction with interconnected pores. Although several experimental and modeling efforts have improved our understanding of tritium transport, the exact mechanisms governing tritium release from breeder materials are still unknown. As a result, current models cannot reliably assess the tritium breeding capabilities of cellular breeder materials. This work presents a tritium transport multiphysics model and calibrates it using experimental data of deuterium absorption and desorption from cellular breeder materials at different temperatures and pressures. This model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. After calibrating the model and performing sensitivity analysis, we discuss pre-dominant mechanisms governing tritium release from cellular breeder materials. The model is part of the Tritium Migration Analysis Program [TMAP8], a multiscale, multiphysics framework for tritium transport based on the finite element multiphysics framework MOOSE. This study demonstrates some of TMAP8’s capabilities and provides insight into the mechanisms governing tritium transport in cellular ceramic breeder materials.

08 HYDROGEN↗
DOE OSTI2025

Corrosion of Wrought and LPBF 316H in Molten Eutectic Na-Mg Chloride Salt

This presentation shows experimental work at INL evaluating the corrosion behavior of wrought and laser powder bed fusion (LPBF) 316H stainless steel (SS) in chloride molten salt systems. This evaluation focuses on understanding the relationship between the material’s as-built characteristics and its corrosion performance, the effects of acid cleaning and descaling before corrosion.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Fusion Energy Research at Idaho National Laboratory: Experimentation and Simulation to Support Safety and Rapid Technology Development

Research into fusion energy is growing rapidly, responding to a call for sustainable sources of energy to replace fossil fuels and mitigate climate change. Within the United States, at least, researchers are also responding to the “Bold Decadal Vision” proposed by the White House, seeking to have a commercially relevant fusion pilot plant deployed within a decade. Before this can become a reality, many Fusion Science & Technology (FS&T) gaps remain. For over 45 years, Idaho National Laboratory has been at the forefront of addressing these FS&T gaps in the context of fusion safety and technology via the operation of world-leading experimental facilities within the Safety and Tritium Applied Research (STAR) Facility. Here, INL focuses on the tritium fuel cycle, conceptual system design studies, risk assessment, waste management, and materials safety. Modeling and Simulation (M&S) has also been a component of this portfolio of research, but, early on, focused on individual systems. Since 2019, active development and research on integrated whole device modeling tools based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework has been undertaken. This has culminated in a MOOSE-based version of the Tritium Migration and Analysis Program (TMAP), an INL code historically focused on tritium permeation and trapping within fusion systems. More recently, INL Laboratory Directed Research and Development funds have been used to create the Fusion ENergy Integrated multiphys-X (FENIX) code focused on scrape-off layer plasma physics and the first wall of a magnetically confined fusion device. This talk will focus on an overview of INL activities in the FS&T research area, with a particular focus on recent M&S activities and results.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Richard P. Feynman Center for Innovation: 2025 Annual Report

The Richard P. Feynman Center for Innovation helps businesses and organizations work with Los Alamos National Laboratory to turn research into practical products and solutions. We connect industry partners with new technologies, research expertise, and licensing opportunities. From protecting inventions to building collaborations, the Feynman Center makes it easier to bring scientific advances to the marketplace and create lasting impact.

08 HYDROGEN↗