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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Flexible Fully-Decoupled Nuclear Plants with Thermal Energy Storage - Technoeconomic Optimization

New nuclear power plants may be one of the technologies for future zero carbon electricity production. These future systems will however also need significant flexibility to cope with fluctuating demand and large share of intermittent renewable supply, rather than running in conventional baseload generation. The present work explores the use of thermal energy storage (TES) as a buffer between the reactor heat and its conversion to electricity, enabling flexible operation of nuclear plants. The focus is on maximizing flexibility, ideally as full decoupling of the nuclear and power cycle islands, as well as on system efficiency in various modes. A system design presented for a high temperature gas cooled reactor (HTGR) presents a slightly complex solution, as separate high- and low-temperature TES systems are required in order to cover the extended temperature range. Although this configuration imposes certain limitations on the discharge system configuration, it also presents the opportunity to design a more efficient system. For this case in particular, a steam reheat cycle is advantageous. Dynamic models were developed for the decoupled system in order to generate insights into aspects of off design operation and system control. A technoeconomic analysis, furthermore, provides sizing and costing of the component, followed by size and dispatch economic optimization considering energy arbitrage. This optimization shows that in all explored markets, the addition of TES system with the proposed configuration results in positive impact on project net present value. Further sensitivity analyses show impact of multiple inputs and possible technical limitations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

HFIR High Power HEU Neutronics Analyses

Department of Energy National Nuclear Security Administration Office of Material Management and Minimization’s mission includes the conversion of civilian research reactors from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. Analyses have shown that the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) will need to operate at 95 MW for the LEU silicide dispersion fuel designs to match key performance metrics obtained with HEU fuel at 85 MW. To prove safe operation of HFIR after installation of plant modifications to increase power, a high power HEU test cycle was proposed. Neutronics model updates and reactor physics analyses are performed to support the development of safety design reports for the high power (HP) HEU test cycle. Reactor physics metrics evaluated herein include fuel depletion, actinide production, cycle length, fission rate density distributions, axial power peaking factors, and reactor kinetic parameters. These reactor physics analyses support the development of future LEU safety design reports by providing key input for future HP HEU HFIR thermal hydraulics and reactor transient safety analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Versatile Heat Transfer Module

Development of micro-reactors that can be easily transported to remote areas for civilian or military applications is an immediate focus of U.S. nuclear industry and therefore is being pursued by multiple reactor vendors (e.g. HolosGen LLC, Westinghouse, Ultra Safe Nuclear Corporation, URENCO, etc.) and actively supported by DOE-NE, DOE-ARPA-E and by the Department Of Defense (DoD). Very-high-temperature gas-cooled (VHTR) and heat pipe cooled reactor technologies are regarded as promising concepts for microreactors applications since they enable compact power conversion systems and display beneficial inherent safety features. Development of microreactors with enhanced performance must be supported by innovations in nuclear technologies. Several technologies being developed by the microreactor industry are being revisited with material technologies developed at the Argonne National Laboratory (ANL). Innovative high-temperature compatible solutions are developed to enable advanced thermal-spectrum microreactor designs featuring improved neutronics performance at high-temperature operating conditions to enable high thermal efficiency and economic competitiveness. This report focuses on the Versatile Heat Transfer Module (VHTM) technology that is developed at ANL for heat-transfer in thermal-spectrum microreactors. This technology is built on a liquid alkali metal heat pipe concept while leveraging advanced nuclear material technologies for high-temperature applications at reduced neutronic penalty with potential improvement in inherent safety behavior. The VHTM can be applied to design high-temperature thermal-spectrum microreactor systems with reduced size, weight, uranium enrichment, or increased lifetime. The INFINIT technology deriving from the VHTM enables design of a highly compact long-life microreactor concept for fission surface power (FSP). Demonstration of multiphysics modeling capability was also completed to showcase the added inherent safety benefits obtained with this concept. The VHTM technology remain at an early development stage with low Technology Readiness Level and Manufacturing Readiness Level. The current demonstration of these technologies has focused on manufacturing capability, on performance of the hydrogen diffusion barrier under thermal cycling, while radiation tolerance demonstration is planned. The next step of the demonstration plan will be the assembly of miniature VHTM for high-temperature testing, which can be used as a prototype for future pilot scale demonstration. Such research is critical to pursue development of these technologies that have the clear potential to significantly improve the performance of currently investigated micro-reactors systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Evaluation of mitigation strategies for radioactive fission gases in fluoride-salt-cooled high-temperature reactors

Here, the Fluoride-salt-cooled High-temperature Reactor (FHR) is a promising Generation IV nuclear reactor due to its improved safety design, near atmospheric working pressure, and high electric power conversion efficiency. However, control of radioactive fission gases, such as tritium, is a critical issue in FHRs due to the significantly larger production rate and potentially larger leakage rate of tritium compared to those in Light Water Reactors (LWRs). Therefore, this study proposes two mitigation strategies for tritium in FHRs: Double-Wall Heat eXchanger with a Tritium Carrier (DWHX-TC) option and Single-Wall Heat eXchanger with a Tritium Barrier (SWHX-TB) option. For the DWHX-TC option, two tube surface configurations, i.e., plain and fluted tubes, and four tritium carriers, i.e., helium, FLiBe, FLiNaK, and KF-ZrF 4 , are initially considered, among which the best tube configuration and promising tritium carrier are then selected and used for design optimization of the DWHX. For the SWHX-TB option, two cases, tritium barriers with and without cracks, are investigated in this study. In addition, the two mitigation strategies are evaluated by an in-house, one-dimensional, coupled HEat and MAss Transfer (HEMAT) code developed to predict the leakage rate of tritium in nuclear systems. Our analysis shows that (1) a double-wall configuration using fluted tubes as both the inner and outer tubes is better than other double-wall configurations investigated; (2) helium, as a tritium carrier, is superior to other candidates, including FLiBe, FLiNaK, and KF-ZrF 4 ; (3) the leakage rate of tritium (into the atmosphere) is reduced from 3400 Ci/day, taking the Advanced High-Temperature Reactor (AHTR) as a reference, to 4.0 Ci/day for an optimum design of the DWHX-TC option using helium as the tritium carrier and 1.7 Ci/day for an optimum design of the SWHX-TB option using 30-µm thick silicon carbide (SiC) as the tritium barrier (no cracks); and (4) potential cracks in tritium barriers may significantly deteriorate mass transfer performance of the SWHX-TB option. For example, a cracking rate of 0.1% leads to an increase of 694.1% in the leakage rate under the conditions investigated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

High Temperature Gas-cooled Reactors: Core Design

General Attributes of Modular Prismatic and Pebble Bed HTGRs Common features and physics Neutronics Thermal-Fluidics Plant Systems and Power Conversion Instrumentation and Control Normal Operation and Power Maneuvers

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Impact of U-10Mo HALEU fuel element tolerances on the Massachusetts Institute of Technology reactor safety and operational performance – Thermal hydraulics

The U.S. is coordinating efforts for the conversion of six U.S. High Performance Research Reactors (USHPRR) including one critical facility from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts of Institute of Technology Reactor (MITR), and achieve similar performance, high-assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being evaluated. The impact of the fabrication specification and tolerances was assessed following the preliminary design of the MITR LEU fuel elements using the U-10Mo monolithic alloy. This research focuses on the analysis of fabrication specification impact on thermal hydraulics (TH) characteristic of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., coolant channel gap thickness, fuel plate thickness, etc.). The analyses are performed based on an all-fresh LEU fuel conversion plan identified in a preliminary safety analysis report submitted to the Nuclear Regulatory Commission. The reactor power margin to the onset of nucleate boiling (ONB) is assessed under the limiting safety system settings (LSSS), where a scram occurs, to ensure there is sufficient margin to the reactor safety limit, which is defined by the onset of flow instability that occurs after the ONB. The best estimate plus uncertainty approach is employed to analyze this TH characteristic, which yields realistic results while maintaining adequate conservatism, utilizing a statistical uncertainty propagation method with the STAT7 code. The TH characteristic is analyzed as a function of the variability of the specification parameters resulting from the fabrication process. The main findings of this study show that the MITR core can meet the TH safety and operational requirements at the all-LEU initial core startup (cycle 1), selected transition cycles (most reactive cycle and most limiting cycle: cycle 3 and 5, respectively) and equilibrium (cycle 14) cores under all limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core power margin to ONB on those specification parameters that have the most direct impact on TH performance is non-linear but monotonically decreasing within the specification tolerances. The third order polynomial fit curves are reported in detail for selected limiting cases and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.

Conversion↗
DOE OSTI2023

Signal Processing of Multiplexed Optical PWM Signals for Sensor Arrays in Nuclear Environments

Safe and effective generation of terrestrial nuclear power greatly benefits from the actionable data provided by the array of sensors located throughout a plant to provide a holistic online indication of reactor operation. This array includes sensors for monitoring coolant flow and pressure, temperature and heat transfer, radiation levels, structure health monitoring, and other critical parameters for reactor operation. While sensors capable of measuring these parameters have been developed, the electronics used in the pre-amplification and analog-to-digital conversion of the small signals they produce are extremely sensitive and susceptible to damage by high-temperatures and radiation environments nuclear reactors encounter while generating power. The small signals from sensors in nuclear power plants (NPPs) are transmitted over long cable runs which introduce dispersion artifacts into the signals of interest as well as electromagnetic interference (EMI) from lighting fixtures, pumps, mains electricity, and other equipment. To overcome these challenges, a front-end digitization (FREND) platform has been developed to use radiation-tolerant electronics to multiplex, amplify, and optically encode signals from an array of sensors for transmission over an optical fiber to mitigate dispersion and EMI artifacts from long runs of electrical cabling. To recover the optically transmitted data, a signal processing scheme based on 1-dimensional template matching to an indexing channel is described herein and demonstrated to have an effective bit-depth of 9.2 bits (1%). This scheme has been validated in proof-of-concept, non-nuclear testing and preliminary experimental results show good agreement between the measured optical output and and sensor input signals. The FREND system represents a low-loss data link between sensors in nuclear environments and data acquisition hardware which is aimed at improving the signal-to-noise ratio of the data acquired from these sensors to provide better information to operators and researchers.

Sweeney, Dan↗
DOE OSTI2025

Laser-based conversion electron Mössbauer spectroscopy of 229 ThO 2

The exceptionally low-energy 229 Th nuclear isomeric state is expected to provide several new and powerful applications, including the construction of a robust and portable solid-state nuclear clock, perhaps contributing to a redefinition of the second, exploration of nuclear superradiance and tests of fundamental physics. Further, analogous to the capabilities of traditional Mössbauer spectroscopy, the sensitivity of the nucleus to its environment can be used to realize laser Mössbauer spectroscopy and, with it, new types of strain and temperature sensors and a new probe of the solid-state environment, all with excellent sensitivity. However, current models for examining the nuclear transition in a solid require the use of a high-bandgap, vacuum ultraviolet (VUV) transmissive host, severely limiting the applicability of these techniques. Here we report the first, to the authors’ knowledge, demonstration of laser-induced conversion electron Mössbauer spectroscopy (CEMS) of the 229 Th isomer in a thin ThO 2 sample whose bandgap (approximately 6 eV) is considerably smaller than the nuclear isomeric state energy (8.4 eV). Unlike fluorescence spectroscopy of the 229 Th isomeric transition, this technique is compatible with materials whose bandgap is less than the nuclear transition energy, opening a wider class of systems to study and the potential of a conversion-electron-based nuclear clock.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Process Heat for Chemical Industry

Process heating with nuclear energy can reduce greenhouse gas emissions by reducing combustion of fossil fuels in fired heaters as steam boilers. Light water reactors can replace the majority of steam duties used by industry; however, high temperature processes such as steam methane reforming require advanced high temperature reactors. Here, guidance on matching the scale of nuclear reactors with specific industries is provided. Principles of heat transport, temperature boosting, and substitution for hot combustion gases are also discussed in this section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Design, modeling and simulation of nuclear-powered integrated energy systems with cascaded heating applications

Nuclear-renewable integrated energy systems (IES) consist of a variety of energy generation and conversion technologies and can be used to meet heterogeneous end uses (e.g., electricity, heat, and cooling demands). In addition to supply-demand balance, end-use heat demands usually require heat supply of certain temperature ranges. The effective and efficient utilization of heat produced within an IES is, therefore, a critical challenge. Here, this paper examines design options of an IES that includes heating processes of multiple temperature grades. We investigate a cascaded design configuration, where the remaining residual heat after high-grade heating processes [e.g., hydrogen production through high-temperature steam electrolysis (HTSE)] is recovered to meet the low-grade heating needs [e.g., district heating (DH)]. Additionally, a thermal energy storage system is integrated into the DH system to address the imbalance between heat supply and demand. This paper primarily focuses on the design and modeling of the proposed system and evaluates its operation with a 24-h transient process simulation using a DH demand profile with hourly resolution. The results indicate that the residual heat from the HTSE exhaust is insufficient for the DH demand, and additional topping heat directly from the reactor process steam is needed. Furthermore, the inclusion of thermal energy storage within the DH system provides the necessary balance between thermal generation and demand, thereby ensuring a consistent rated temperature of the DH supply water. This approach helps minimize the control actions needed on the reactor side.

08 HYDROGEN↗
DOE OSTI2023

AI-based Detection and Defense Against Cyberattacks in Distributed Energy Resources

This study will provide comprehensive artificial intelligence (AI)-based solution tools for network security, malware prevention, and sensor data anomaly detection for distributed energy resource (DER) research, development, and demonstration. DER technologies are energy systems (e.g., solar panels, wind turbines, and energy storage systems) that are often connected to the internet and thus vulnerable to cyberattacks. Cybersecurity should be of primary concern for DERs, which is why we propose an integrated multi-layer cyber-defense system for DERs. This system encompasses risk assessments, network security, malware prevention, and detection of anomalies in the sensor data. Implementation of a comprehensive risk assessment with an overview of the model architecture should be the primary step, and should include the potential impact of experiencing, at a given time, one or more cyberattacks on the system. The second step is to ensure that the network security includes firewalls, intrusion detection, and malware prevention. The third step is to provide solution tools that enable sensor data anomaly detection for DERs. By incorporating these considerations into DER research, development, and demonstration, organizations can help ensure the safety and security of their systems and protect against potential cyberattacks.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2026

Modeling and analysis of synthetic liquid fuel production from CO 2 and nuclear energy using methanol-to-diesel process

Electrofuels (e-fuels) are synthetic fuels produced from carbon dioxide (CO 2 ) and electricity for blending with or replacing petroleum fuels. Nuclear energy is an attractive energy feedstock for e-fuel production because of its low environmental footprint and its ability to provide steady heat and power essential for e-fuels production. We modeled and evaluated the cost and environmental footprint of e-fuels production in the distillate range for three nuclear power scales, 100, 500, and 1000 MWe, through methanol and olefins intermediates leveraging commercial or high technology readiness level (TRL) processes. Compared to the commonly studied e-fuels from Fischer Tropsch process that has a distillate yield of <70% with the rest being low value naphtha, the proposed process via methanol intermediate increases the product selectivity with distillate yield of 96% and only 4% naphtha. The modeled process has a carbon conversion ratio of 98%, and a process energy efficiency of 56% relative to the total equivalent nuclear electricity input. The e-fuel plant economics and GHG emissions were estimated by considering CO 2 collected from ethanol plants adjacent to nuclear power plants. The estimated minimum fuel selling prices (MFSP) of e-fuel is in the range of $5.7-$9.1/gal depending on e-fuel plant scale, electricity cost, and CO 2 transportation distance. The corresponding e-fuels life cycle GHG emissions is estimated in the range of 5-6 gCO 2 e/MJ of liquid fuel using the R&D Greenhouse gases, Regulated Emissions, and Energy use in Technologies (R&D GREET) model.

10 SYNTHETIC FUELS↗
DOE OSTI2024

Design and Prototyping of Advanced Control Systems for Advanced Reactors Operating in the Future Electric Grid (Final Report)

Despite its significant advantages as a baseload, low-carbon energy source, the U.S. nuclear power industry has faced increasing difficulties in maintaining economic competitiveness in a rapidly evolving energy market. The economic conditions faced by the current fleet of nuclear power plants (NPPs) in the U.S. deregulated electricity market require a concerted effort to mitigate specific cost factors. Many units are struggling to stay competitive, and some premature shutdowns have occurred. Besides, in response to the large penetration of renewable energy sources, the role of nuclear power plants as pure baseload units needs to be reconsidered. Based on these experiences, operational flexibility is considered a fundamental requirement for the next generation of nuclear reactors to be competitive in the future energy market. The deployment of advanced reactors capable of operating within a new power grid paradigm, known as the Integrated Energy System (IES) was investigated. This approach combines new reactor designs with Thermal Energy Storage (TES) technologies, allowing the nuclear reactor to maintain a steady power output without the need for constant adjustments in response to load demand fluctuations. With this configuration, the reactor operates as a baseload unit, experiencing only very gradual power transients, while the energy storage facility within the power conversion cycle acts as a peaking unit.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Alternative Power and Propulsion for Space Systems

What are the main sources of power and propulsion for space systems and what are their advantages, disadvantages, how do they work, and what are the politics behind them? The following topics will be analyzed solar panels, batteries, radioisotope thermoelectric generator (RTG), advanced Stirling radioscopic generator (ASRG), reactors, nuclear thermal rockets (NTP), thermionic experiment with conversion (TOPAZ), electrothermal plasma thruster (VASIMR), ion propulsion, and finally nuclear pulse propulsion. The production of Pu-238 is extremely controversial but a vital source of fuel for many of these space systems. Soral and Batteries are the most common and the least controversial forms of power used in space systems today. However, is ineffective for deep space exploration as they only function for up to four years without needing to be recharged. This is where nuclear systems come into play. Nuclear systems such as an RTG will “live” up to 88 years. For example, Voyager 1 is powered by an RTG It has been traveling since 1977 and has left out solar system in 2012 and has yet to run out of power. However, this type of technology (nuclear) has many hurtles it needs to jump through before advancements in can go much further.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Sandia Nuclear Cartridge Concept

Nuclear power offers the promise of long-term electrical power for remote areas. Recent advances in passive safety and long-life cores make a reactor that can be operated autonomously for 20 years or more a real possibility. This white paper discusses a reactor concept that offers the potential for further development, resulting in a permanently hermetically-sealed "nuclear cartridge." The term "nuclear cartridge" is meant to imply a nuclear energy source that can be inserted into a site and operated autonomously until its energy has been depleted, then withdrawn and replaced by another cartridge. The concept can be scaled for various sizes, ranging from about 1 megawatt-electric (MWe) to about 100 MWe. The paper also discusses the concept of Integrated Safety, Operations, Security, and Safeguards (ISOSS) by design as it applies to this reactor design. Finally, a discussion of smart grids and how they can benefit the transfer of power to the end user is included. The Nuclear Cartridge concept has been developed with the following characteristics in mind: highly reliable autonomous operation coupled with international monitoring, requiring minimal on-site operations personnel; walkaway passively safe design; cartridge replacement cycle on the order of 20 years; load following capability; physical security by design requiring minimal security personnel during operations; and proliferation resistance by design. As illustrated in figure 1, integrating the reactor with advanced power conversion, smart grids, and other sources of energy results in a resilient and sustainable energy source.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Thermal Hydraulic Modeling of an Advanced Nuclear Reactor using open-source MOOSE tools

The aim of this study is to develop model of a small High Temperature Gas cooled Reactor (HTGR) including the balance of plant. This microreactor produces electricity using the thermal power of the nuclear reaction. This work utilizes the MOOSE Multiphysics simulation tools, which are mainly developed at Idaho National Laboratory (INL, Idaho, United States of America). Its thermal-hydraulics and Heat conduction Modules are used to study the fluids behavior in the primary loop and power conversion system and their interactions with the heating structures. More specifically, one verifies that the temperatures, pressures and mass flow rates of the fluids in both loops are consistent and that, at the same time, all the power transfers occur as expected. Moreover, the various mechanical components characteristics (turbine, compressor, pump) are adapted to the reactor operating conditions. This first analysis is conducted using simplified and one-dimensional model for the core. Ultimately, the obtained results are used to build a higher fidelity core model. This step aims to verify that the initial simplified simulation is consistent with the three-dimensional core modeling. In addition, the calculated material temperatures are checked. This study provides a fairly complete model of the thermal hydraulic phenomena for a particular design of High Temperature Gas cooled Reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Nuclear-Driven Chemical Conversion Pathways

The purpose of this work is to explore the possibilities of using coal as a carbon resource by converting it to fuels and chemicals with heat from a nuclear power plant. While many previous studies have focused on coal gasification, this process requires combusting a portion of the coal and producing heat. By employing pyrolysis over gasification, the energy content in the coal can be preserved. Pyrolysis of coal also results in a large portion of solid char, which can be converted to activated carbon and sold to various markets. This model uses hydrogen produced by nuclear-supplied heat and electricity to convert the syngas stream to methanol, which has many pathways for further processing into fuels, chemicals, and plastics. Specifically, this project focuses on non-fuel products for methanol pathways. The CO2 captured within the refinery model can also be combined with hydrogen to produce more chemical products and reduce CO2 emissions.

01 COAL, LIGNITE, AND PEAT↗