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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Select Proliferation Studies on TRISO-fueled, Heat-Pipe-cooled Microreactors

Nuclear microreactors carry the potential to open up new markets for the nuclear industry, as their expected cost competitiveness in non-traditional market segments (e.g., mines, military bases, extraterrestrial surfaces, and remote areas), and their inherent safety features make them deployable when other power sources are unavailable or difficult to exploit. For countries that have not traditionally participated in nuclear power, microreactors represent a clean energy solution [1]. However, their use, especially in non-weapons states, may entail challenges in terms of maintaining international nuclear safeguards [2]. Furthermore, the deployment locations where microreactors may prove most cost competitive would be difficult to access by state and International Atomic Energy Agency (IAEA) inspectors [3]. In addition to the isolated nature of potential deployment sites, the low-power characteristic of microreactors suggests that numerous microreactors would need to be deployed to meet energy demands. That, coupled with the unique physics of many current microreactor designs, opens up a new area of research with respect to nonproliferation and safeguards concerns [2]. Whereas traditional facilities are inspected as isolated cases when looking for signs of diversion or misuse; microreactors may need to be assessed in the context of the entire fleet to which they belong. International safeguards necessitate timely detection of any significant quantities (SQs) of material that are being diverted (e.g., 1 SQ of special nuclear material diverted over the course of a 1-year period) [4]. For low-enriched uranium, the IAEA defines 1 SQ as corresponding to 75 kg of 235U. The purpose of the present paper is to explore the detectability threshold for material diversion in microreactors by relying on critical control drum angles, excess reactivity, and the reactor lifetime as the selected operational parameters. For this assessment, a heat-pipe-cooled microreactor was regarded as the base design. While the conclusions reached in this paper are not readily extendable to the design of actual microreactors, the analysis herein enables conclusions to be drawn regarding the level of accuracy needed for reference calculations in order to detect diversion scenarios by utilizing the selected operational parameters (i.e., mainly control drum angles, critical insertion angle, and the reactor lifetime).

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

A Very High Temperature Lead Cooled Microreactor Concept using Advanced Materials and Thermophotovoltaics Direct Energy Conversion Technology

Recently, significant advances have been made in designing high-temperature capable nuclear fuels. In parallel, important advances have occurred in design of nanophotonic structures (coatings) with ability to shape the frequencies of light being emitted from hot surfaces. In this project, we bring these materials advances together into a design for a new direct conversion nuclear powered system with high thermodynamic efficiency, inherent safety, and unprecedented reduction in size and weight relative to current designs. The concept involves using the heat generated from nuclear fission to raise the temperature of a selective thermal emitter to >800°C. The selective emitter shifts the normal broad brand emissions of light from its hot surface into the correct near- and mid-IR bands that produce electricity with inexpensive photovoltaic cells. We have designed a new annular flow Lead Fast Reactor (LFR) that minimizes overall reactor/power conversion footprint. The results indicate that the overall microreactor/power-system design offers a 2X reduction in size over existing technology that is based on the sCO 2 reverse compression Brayton cycle. Additionally, the only moving part in our design is a circulation pump to cool the PVT panels.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Decay heat analysis for advanced reactor spent fuel transportation and storage applications

Accurate characterization of nuclide inventories and decay heat in spent nuclear fuel is critical for ensuring its safe handling, storage, transportation, and disposal. Although extensive research has been conducted on light-water reactor fuel, advanced reactors present unique challenges due to their diverse core configurations, fuel characteristics, neutron energy spectra, and burnup levels. Building upon previous efforts that developed representative reactor core models for various advanced reactor types and fuels, this study evaluates reactor-specific decay heat characteristics. The results highlight significant variations across advanced reactor types as well as across reactor designs within the same reactor type, and they provide comparison to typical commercial light-water reactor fuel. For example, thermal-spectrum reactor fuels were observed to have an approximately 100-fold decrease in decay heat over the first decade of cooling, whereas the reduction was 10-fold for fast-spectrum reactor fuels. Mass-specific decay heat at discharge can differ by three orders of magnitude among the fast and thermal reactor systems considered. Overall, for the analyzed advanced reactor fuel, fewer than 17 nuclides account for over 99% of total decay heat at 0.5 years, and that number drops to fewer than 7 nuclides at 100 years of cooling. By quantifying reactor-specific decay heat trends and nuclide contributions, this work provides a technical basis to support the development of spent fuel management strategies for advanced reactor fuels as well as safety evaluations for storage, transportation, and long-term waste disposal.

Advanced reactors↗
DOE OSTI2020

Mechanochemical Synthesis of High-Purity Anhydrous Binary Alkali and Alkaline Earth Chloride Mixtures

A direct synthesis route for high-purity, anhydrous binary salt mixtures has been developed. This atom efficient, solvent-free process is easily scalable, with the potential to produce salt mixtures that meet the purity standards required for industrial heat transfer and nuclear applications. The essence of the methodology lies in mechanochemical synthesis of carnallite precursors that can mitigate the hydrolysis of MgCl 2 ·6H 2 O under direct heating. Each dehydrated salt carnallite was then analyzed for purity and oxide content through subsequent powder X-ray diffraction, and strong acid titration. In conclusion, this process presents a more effective alternative route compared to previous methods for obtaining low-oxide, high-purity chloride salt mixtures.

Alkali↗
DOE OSTI2021

SCALE 6.2.4 Validation for Light Water Reactor Decay Heat Analysis

Energy release from the decay of radionuclides in nuclear fuel after its discharge from reactor is a critical parameter for design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. Well-validated computational tools and nuclear data are essential for decay heat prediction. This paper summarizes the validation of the SCALE nuclear analysis code system version 6.2.4, used with ENDF/B-VII.1 libraries, for decay heat analysis of light water reactor used fuel. The experimental data used for validation include full-assembly decay heat measurements that cover assembly burnups of 5 to 51 GWd/tonne U, cooling times after discharge in the 2- to 27-year range, and initial fuel enrichments up to 4 wt% 235U. The comparison between calculated (C) and experimental (E) decay heat showed very good agreement, with an average C/E over all considered measurements of 1.006 (σ = 0.016) for pressurized water reactor and 0.984 (σ = 0.077) for boiling water reactor assembly measurements. The effect of using assembly-average versus axially varying modeling data on the calculated decay heat, important to thermal analyses for used fuel transportation and storage systems, is discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Coupled hydro-thermal flow and radionuclide transport driven by spatial variation of heat-generating radioactive wastes in shale formations

Deep geologic disposal of multiple nuclear waste packages with various heat sources can induce nonuniform hydro-thermal behaviors in the near-field of the repository, consequently influencing the long-term radionuclide transport in the far-field once waste form breach initiates. Here, this study looks into three cases with variation in the spatial order of six groups of heat sources (10th, 50th, 75th, 90th, 95th, and 99th percentiles of heat outputs generated from 1,981 as-loaded dual-purpose canisters in the field site) in a shale-hosted repository with respect to the uni-directional groundwater flow (from west to east): (1) cooler waste packages from west to east, (2) hotter waste packages from west to east, and (3) hottest waste packages in the middle of the repository. Our field-scale PFLOTRAN simulation represents heat-driven multiphysics coupled mechanisms, including multiphase flow, heat transfer, and chemical/radioactive transport, and also, calculates the onset of waste form breach based on temperature-dependent canister vitality. The results from this sensitivity study will quantify the short- (less than 1 × 10 3 years) and long-term (up to 1 × 10 6 years) impacts of sporadic heat pulses from waste package on the spatio-temporal perturbation in hydro-thermal flow quantities and the rate of radionuclide transport in both near- and far-field of the repository system.

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

Advanced Reactors Integrated Energy System: Thermal Energy Storage Island Design

The main topic of this research is integrated energy systems (IES) designed for pairing industrial thermal energy loads with advanced reactors (ARs). The Idaho National Laboratory (INL) Crosscutting Technology Development IES program and the National Reactor Innovation Center (NRIC) are seeking to develop, design, and construct an AR-IES demonstration platform that couples the thermal output from an AR operating at the INL/NRIC Demonstration of Microreactor Experiments (DOME) test bed in the Experimental Breeder II dome to a variable capacity load emulator (i.e., air-cooled radiator) and sensible thermal energy storage (TES) via a molten salt thermal energy transfer fluid. In the rapidly evolving landscape of energy supply and distribution, flexibility has emerged as a prized attribute, surpassing the traditional notions of stability and baseload generation capability. This shift in priorities is particularly evident in the context of nuclear power plants (NPPs), where adaptability over constant output is becoming more important. As our energy infrastructure and resources embraces the rise of distributed energy generation, the inherent variability in net demand continues to grow. Moreover, the use of nuclear energy as a source of heat for decarbonizing the industrial sector is becoming a very pressing topic. In such environment, advanced NPPs are poised to enter a more competitive energy market, delivering both, flexible electricity and heat. This shift motivates the exploration of TES systems, designed to empower NPPs with nimble responsiveness to market fluctuations, flexible heat delivery capabilities, and redefine their role in the energy field. TES systems offer the unique advantage of storing nuclear energy in its original form as heat, thereby affording unparalleled flexibility in its subsequent utilization.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2006

Gas Foil Bearings for Space Propulsion Nuclear Electric Power Generation

The choice of power conversion technology is critical in directing the design of a space vehicle for the future NASA mission to Mars. One candidate design consists of a foil bearing supported turbo alternator driven by a helium-xenon gas mixture heated by a nuclear reactor. The system is a closed-loop, meaning there is a constant volume of process fluid that is sealed from the environment. Therefore, foil bearings are proposed due to their ability to use the process gas as a lubricant. As such, the rotor dynamics of a foil bearing supported rotor is an important factor in the eventual design. The current work describes a rotor dynamic analysis to assess the viability of such a system. A brief technology background, assumptions, analyses, and conclusions are discussed in this report. The results indicate that a foil bearing supported turbo alternator is possible, although more work will be needed to gain knowledge about foil bearing behavior in helium-xenon gas.

Howard, Samuel A.↗
NASA NTRS1979

Reactor design for nuclear electric propulsion

The paper analyzes the consequences of heat pipe failures, that resulted in modifications to the basic design of a heat-pipe cooled, fast spectrum nuclear reactor and led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO2 and molybdenum sheets that span the diameter of the core. Design characteristics are presented and compared for two reactors. A conceptual design for a heat exchanger between the core and the thermionic converter assembly is described. This heat exchanger would provide design and fabrication decoupling of these two assemblies.

Koenig, D. R.↗
DOE OSTI2025

Device for steam cladding oxidation testing at TREAT

To compare the chemical degradation of conventional zirconium alloy (Zry) cladding to advance silicon carbide (SiC) cladding in a post loss of coolant accident (LOCA) environment, new nuclear testing capabilities are necessary. The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) has matured its transient fuel testing capabilities since its 2017 restart. The most recent experiment architecture is the Transient Water Irradiation System in TREAT (TWIST), which is designed to support qualification of accident tolerant fuels in light water reactors. INL has designed and analyzed a natural circulation steam flow modification for TWIST to produce prototypic conditions of cladding oxidation. The in-situ device will be electrically heated to drive natural circulation. Moreover, the SiC cladding requires heating above 1700 °C to observe failure, thus internal prototypic nuclear heating with radiation effects will be used. Thermal hydraulic analysis with RELAP5-3D (Reactor Excursion and Leak Analysis Program) estimated steam fluxes greater than 50 mg cm −2 s −1 can be achieved. These fluxes are adequate to test Zry cladding according to draft regulatory guides and to test SiC cladding according to past experiments.

Oxidation↗
NASA NTRS1981

Heat pipe heat rejection system and demonstration model for the nuclear electric propulsion (NEP) spacecraft

The critical evaluation and subsequent redesign of the power conversion subsystem of the spacecraft are covered. As part of that evaluation and redesign, prototype heat pipe components for the heat rejection system were designed fabricated and tested. Based on the results of these tests in conjunction with changing mission requirements and changing energy conversion devices, new system designs were investigated. The initial evaluation and redesign was based on state-of-the-art fabrication and assembly techniques for high temperature liquid metal heat pipes and energy conversion devices. The hardware evaluation demonstrated the validity of several complicated heat pipe geometries and wick structures, including an annular-to-circular transition, bends in the heat pipe, long heat pipe condensers and arterial wicks. Additionally, a heat pipe computer model was developed which describes the end point temperature profile of long radiator heat pipes to within several degrees celsius.

Ernst, D. M.↗
NASA NTRS1971

Radiation heat transfer calculations for the uranium fuel-containment region of the nuclear light bulb engine.

Calculation results are reviewed of the radiant heat transfer characteristics in the fuel and buffer gas regions of a nuclear light bulb engine based on the transfer of energy by thermal radiation from gaseous uranium fuel in a neon vortex, through an internally cooled transparent wall, to seeded hydrogen propellant. The results indicate that the fraction of UV energy incident on the transparent walls increases with increasing power level. For the reference engine power level of 4600 megw, it is necessary to employ space radiators to reject the UV radiated energy absorbed by the transparent walls. This UV energy can be blocked by employing nitric oxide and oxygen seed gases in the fuel and buffer gas regions. However, this results in increased UV absorption in the buffer gas which also requires space radiators to reject the heat load.

Rodgers, R. J.↗
NASA NTRSDate not supplied

NTAC Augmented Nuclear and Advanced Propulsion

Currently, there are two types of nuclear propulsion: Nuclear Thermal Propulsion (NTP) and Nuclear Electric Propulsion (NEP). Common to both types, a nuclear reactor generates heat energy to raise the temperature of propellant, or the generated heat energy is converted into electric power. The electricity generated powers propulsion systems, such as a Hall thruster, magnetoplasma dynamic (MPD) thruster, or variable specific impulse magnetoplasma rocket (VASIMR). Customarily, only the thermal energy output from nuclear reaction is considered for NTP and NEP systems. However, the nuclear fission reaction can also generate ~7 MeV of prompt gamma rays. Therefore, most designers of NTP and NEP systems incorporate safety features to shield against this radiation. An optimized design approach would be to integrate a Nuclear Thermionic Avalanche Cell (NTAC) within the NTP or NEP structure for converting this radiation energy into additional electric power, while also shielding gamma rays. Consequently, this study addresses several combinations of NTP and NEP structures with NTAC devices in order to achieve that goal.

NTAC↗
DOE OSTI2023

Triply Periodic Minimal Surfaces Lattice Neutronics Scoping Studies

Triply Periodic Minimal Surfaces (TPMS) have seen extensive research in heat exchanger development due to their enhanced heat transfer performance. When applied to nuclear reactor cores, TPMS can more efficiently dissipate heat to the coolant, permitting increased power densities. The increased power density can be used to develop smaller reactors while retaining power output. Although TPMS core geometries promise enhanced heat transfer, the neutronic behavior of such geometries require further inquiry. In this study, the neutron multiplication and buckling are compared between variations of TPMS and traditional pincell core designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Multi-objective Decisions on Integrated Energy Systems Planning and Operation for Industrial Combined Heat and Power Supply

Unlike the power sector—which can transmit electricity over long distances via established grids—the industrial sector poses a unique challenge due to its geographically concentrated large-scale heat processes. Enhancing energy security in such industrial parks provides a dual benefit: reduced exposure to volatile fossil fuel prices and improved economic viability, largely driven by economies of scale in energy supply and distribution. This study presents a comprehensive technoeconomic analysis of a nuclear energy hub, employing a mechanism-focused approach to evaluate uncertainties in operational strategies and capacity optimization. Load profiles from three major energy intesive industries—chemical, refinery, and steelmaking—are examined, each presenting unique challenges and opportunities for nuclear energy integration. We adopt a multi-objective optimization framework, converting multiple objectives into a single objective function through the e-constraint method. The findings highlight clear trade-offs between system conditions and varying levels of energy independence. Overall, this analysis is granular enough to address industry-specific concerns yet sufficiently generalizable to provide actionable insights into the feasibility of nuclear-based clean heat solutions for the industrial sector.

25 - ENERGY STORAGE↗
NASA NTRSDate not supplied

Sensitivity Analysis of Heat Rejection and Propellant Management Technologies for Nuclear Thermal Propulsion Architectures

Cryogenic fluid management (CFM) technologies are very important for enabling a wider range of missions to utilize space nuclear propulsion (SNP) concepts such as nuclear thermal propulsion (NTP). Technologies for thermal and cryogenic propellant management allow for vehicles to take full advantage of the higher efficiency NTP systems for longer duration human interplanetary and deep space robotic missions. Currently, the CFM Portfolio Project at NASA’s Marshall Space Flight Center (MSFC) is developing thermal and propellant management technologies needed for SNP. When developing these technologies, it is important to understand the sensitivities of key performance parameters (KPPs) at the system and overall mission level due to the ways the technologies interact with each other, other subsystems, and influence the overall vehicle. The Advanced Concepts Office (ACO) at NASA-MSFC was tasked with building an integrated system model of a human Mars NTP mission to evaluate the impacts and sensitivities of CFM technologies on the overall vehicle and mission. This paper will cover the buildup of the model and highlight major sensitivities and breakpoints encountered, as well as future work in improving the existing models and sensitivities being evaluated.

Robert J. Hetterich↗