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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The thermal decomposition of the benzyl radical in a heated micro-reactor. II. Pyrolysis of the tropyl radical

Cycloheptatrienyl (tropyl) radical, C{sub 7}H{sub 7}, was cleanly produced in the gas-phase, entrained in He or Ne carrier gas, and subjected to a set of flash-pyrolysis micro-reactors. The pyrolysis products resulting from C{sub 7}H{sub 7} were detected and identified by vacuum ultraviolet photoionization mass spectrometry. Complementary product identification was provided by infrared absorption spectroscopy. Pyrolysis pressures in the micro-reactor were roughly 200 Torr and residence times were approximately 100 μs. Thermal cracking of tropyl radical begins at 1100 K and the products from pyrolysis of C{sub 7}H{sub 7} are only acetylene and cyclopentadienyl radicals. Tropyl radicals do not isomerize to benzyl radicals at reactor temperatures up to 1600 K. Heating samples of either cycloheptatriene or norbornadiene never produced tropyl (C{sub 7}H{sub 7}) radicals but rather only benzyl (C{sub 6}H{sub 5}CH{sub 2}). The thermal decomposition of benzyl radicals has been reconsidered without participation of tropyl radicals. There are at least three distinct pathways for pyrolysis of benzyl radical: the Benson fragmentation, the methyl-phenyl radical, and the bridgehead norbornadienyl radical. These three pathways account for the majority of the products detected following pyrolysis of all of the isotopomers: C{sub 6}H{sub 5}CH{sub 2}, C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2}. Analysis of the temperature dependence for the pyrolysis of the isotopic species (C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2}) suggests the Benson fragmentation and the norbornadienyl pathways open at reactor temperatures of 1300 K while the methyl-phenyl radical channel becomes active at slightly higher temperatures (1500 K).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Proposed Adaptation of the Standard Review Plan NUREG-0800, Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

This report provides a proposed adaptation of a previous regulatory gap analysis of Chapter 4 (Reactor) of NUREG-0800 , Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, for applicability to advanced reactors; specifically, the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for the review because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail. This report satisfies M3AT-16OR2001042, “Complete Development of an Advanced Reactor Adaptation of Chapter 4 NUREG-0800,” in support of the US Department of Energy (DOE) Office of Nuclear Energy (DOE-NE) Advanced Reactor Technology program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Techno-Economic Analysis of an Intensified Integrated Gasification Combined Cycle (IGCC) Power Plant Featuring a Combined Membrane Reactor - Adsorptive Reactor (MR-AR) System

In this work, the novel concept of a combined membrane-adsorptive reactor sequence (MRAR) is developed and implemented in an Integrated Gasification Combined Cycle (IGCC) plant. This novel MR-AR IGCC plant is subsequently analyzed from an economic viewpoint through a techno-economic analysis (TEA) of the proposed plant. This novel design can achieve over 90% carbon capture without the use of a Dual-Stage Selexol unit. The resultant intensified design is more efficient from both an economic and power production perspective than the traditional IGCC plants with pre-combustion Carbon-Capture Storage (CCS) technology. The COMSOL software package is utilized to simulate the MR-AR sequence proposed in this work, and the UNISIM software (HoneywellTM) is used to create an intensified process flowsheet of the proposed MR-AR IGCC plant, which is subsequently heat-integrated. The TEA developed for the MR-AR IGCC power plant will be used to identify the extent of process intensification the proposed design has over the traditional IGCC plants with pre-combustion CCS. Furthermore, the results demonstrate a reduction in both the cost-of-electricity (COE) and in the capital cost of the proposed design over the baseline case.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Summary Report Of The FY25 Reactor Physics Verification And Validation Exercises In The Advanced Reactor Technologies - Gas-cooled Reactor Program

Valdiation and verification of numerical tools is critical for ensuring reasonable predictions for design scoping, licensing, and safety analsyis. In this report, two reactor physics verification and validation exercises are presented. The first of these exercises focuses on burnup analysis with data from the Advanced Gas Reactor (AGR) program. Simulations are performed with Monte Carlo N-Particle (MCNP) and are compared with the experimental measurements for the AGR 1 and 2 experiments that utilize both UCO and UO2 fuel. The second exercises utilizes data from the HTR-Proteus experiments to perform reactor physics validation. Specifications of the experimental facility are provdied, along with a demonstration of initial modeling efforts in Serpent for one of the determistic packing experiments. Both cases are part of the Generation-IV international forum (GIF) Very High-Temperature Reactor (VHTR) Computational Methods, Validation, and Benchmarking (CMVB) program, an international collaborative organization dedicated to the verification and validation of High-Temperature Gas-Cooled Reactor (HTGR) analysis. Participation in the CMVB allows the US Department of Energy (DOE) to leverage these existing validation activities to provide extra value through benchmarking activities with other CMVB members.

and Benchmarking (CMVB) program↗
DOE OSTI2021

Materials for Small Nuclear Reactors and Micro Reactors, Including Space Reactors

Small nuclear reactors, including small modular reactors (SMRs) or reactors for space applications, rely on different materials than those typically applied in large-scale nuclear power plants. Examples include molten salts as cooling medium or fuel carrier, metal hydrides as high-temperature moderators, and fuels allowing for higher burnup. All of these also require novel structural materials, for which material interactions have to be understood. Fissionable and fissile materials, such as uranium or plutonium, are rarely considered in materials design other than for nuclear fuels. Similarly, the aspects of radiation damage, occurring during irradiation when a reactor operates, are unique to nuclear materials research. The handling of these materials puts further limitations on the materials science conducted for nuclear materials. All of these issues move research for these materials off the “main stream” of materials science, and cause it to be more easily conducted at national laboratories. However, nuclear reactors offer unique opportunities for carbon-neutral energy generation and have great potential to address if not solve problems arising from global warming. This special topic, sponsored by the TMS Nuclear Materials Committee, focuses on materials research for small nuclear reactors, both experimental and simulation/modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Nuclear reactor assemblies, nuclear reactor target assemblies, and nuclear reactor methods

Reactor target assemblies are provided that can include a housing defining a perimeter of at least one volume and Np or Am spheres within the one volume. Reactor assemblies are provided that can include a reactor vessel and a bundle of target assemblies within the reactor vessel, at least one of the target assemblies comprising a housing defining a volume with Np or Am spheres being within the volume. Irradiation methods are also provided that can include irradiating Np or Am spheres within a nuclear reactor, then removing the irradiated spheres from the reactor and treating the irradiated spheres.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. Here, the back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Design of a Combined Proportional Integral Derivative Controller to Regulate the Temperature Inside a High-Temperature Tubular Solar Reactor

Solar fuels are proven to be promising candidates for thermochemical energy storage. However, the transient nature of solar radiation is an obstacle to maintaining a stable operational temperature inside a solar reactor. To overcome this challenge, the temperature of a solar reactor can be regulated by controlling the incoming solar radiation or the feedstock flowrate inside the reactor. In this work, a combined proportional integral derivative (PID) controller is implemented to regulate the temperature inside a high-temperature tubular solar reactor with counter-current flowing gas/particles. The control model incorporates two control systems to regulate incoming solar radiation and gas flow simultaneously. The design of the controller is based on a reduced-order numerical model of a high-temperature tubular solar reactor that is vertically oriented with an upward gas flow and downward particle flow. The reactor receives heat circumferentially through its wall over a finite segment of its length. Formulation of the heat transfer model is presented by applying the energy balance for the reactor tube and considering heat and mass transfer inside. A set of governing differential equations are solved numerically by using the finite volume method to obtain reactor wall, particles, and gas temperatures along the reactor length with various boundary conditions. Simulation results are used to tune the PID controller parameters by utilizing the Ziegler–Nichols tuning method. Both the simulation results and the controller performance are visualized on the LABVIEW platform. The controller is challenged to track different temperature setpoints with different scenarios of transient solar radiation. Here, the performance of the PID controller was compared to experimental results obtained from an industrial PID controller embedded in a 7 kW electric furnace. Results show that the combined PID controller is successful in maintaining a stable temperature inside the reactor by regulating the incoming solar radiation and the flowrate via small steady-state error and reasonable settling time and overshoot.

14 SOLAR ENERGY↗
DOE OSTI2017

Emerging and existing sensing technologies for small modular reactors and advanced reactors - 191

The development of small modular reactors (SMRs) and Generation-IV advanced reactor designs represents a promising future for nuclear energy. These next-generation reactors can support new applications for nuclear energy like hydrogen production, industrial heat generation, and water desalination in addition to electricity generation. Furthermore, SMRs and advanced reactors enable load-following operations, offer improved passive safety features, support lengthened refueling cycles, and provide spent fuel management capabilities. To date, there are no SMRs operating or under construction in the United States, but commercial operation of the first integral Pressurized Water Reactor (iPWR) SMR is expected by the mid-2020's with Generation-IV advanced reactors to follow a decade or two later. To facilitate the development and deployment of SMRs and advanced reactors, key instrumentation and control (I and C) system challenges must be addressed and existing nuclear grade I and C sensors will need to be adapted or new sensors developed. For example, traditional measurement techniques and maintenance strategies used in the current generation of light water reactors (LWRs) will be challenged by the inherent design characteristics of the iPWR. For Generation-IV reactors, conventional I and C sensors will not withstand prolonged exposure to the harsh environmental conditions of advanced reactors (e.g. elevated temperatures, corrosive coolants, increased radiation). This paper discusses the challenges associated with making primary system process measurements in next-generation reactors and includes a summary of work performed to evaluate existing and emerging sensor technologies for these systems. As importantly, the paper presents the initial test results of a prototype temperature sensor developed by Analysis and Measurement Services Corporation (AMS) for use in SMRs and advanced reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Need for research and training reactors for advanced reactor designs

Full text of publication follows. Research and training reactors have served a valuable role in helping train workforce for currently operating fleet of light water reactors. These research and training reactors have been used in reactor laboratory classes to familiarize the students with such vital concepts as approach to criticality, reactor period, neutron moderation, reactivity, flux distribution and leakage, etc. As the industry moves toward advanced non-light-water reactor designs, it is critical that research and training reactors be developed and deployed at university campuses to help train the new generation of nuclear and non-nuclear engineers who are likely to design, build, and operate these advanced reactors. Among the designs currently being pursued for nuclear power generation include molten salt, sodium cooled, and gas cooled designs, with options for various fuel forms. Thus, industry and DOE in collaboration with academic institutions should devise plans on how to familiarize the next generation of nuclear workforce with hands-on experience necessary for such designs. These research and training reactors will play a vital role in familiarizing the future workforce with hands-on experience with concepts associated with fast spectrum reactors, gas cooled reactors, and other features not associated with light water reactors. In addition to classical nuclear engineering concepts, these advanced research and training reactors can also be used for hands-on training as well as for research on features being considered in the design of GEN-IV reactors: cyber security for digital control room operations, hybrid energy system, hydrogen generation, district heating, autonomous control... (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Need for advanced research reactors for the next-generation reactor physics, analysis tools, and technology

Full text of publication follows. There is an urgent need for design and deployment of advanced research and test reactors in support of design, licensing and operation of advanced power reactors and education of next generation nuclear workforce. Existing research reactors mostly were designed and constructed decades ago with the main objectives of training operators, performing reactor physics experiments, and educating nuclear engineers and scientists. There are already gaps and significant concern about future capabilities for the existing research reactor facilities to address modern instrumentation and/or flexible environments for: performing reactor physics studies for advanced designs which have significantly different core materials forms and compositions, reactor shapes and size; validation of advanced high-fidelity software; development of machine learning algorithms for enhancement of human-machine collaboration in support of reactor monitoring, operation and safeguards; and, effective education of the next-generation workforce. The authors will focus on the need for advanced research reactors to improve and validate fast and accurate simulation tools for high-fidelity modeling and analysis of nuclear reactors in support of their design, optimization, licensing, operation, and monitoring. In the past, simulation tools were limited to relatively coarse models using approximate methodologies that benefited from two main factors: i) allowance for large margins and tolerances; ii) ability to construct prototype (e.g., zero power) reactors for adjustment of approximate methodologies. The next generation reactors have to be designed mainly by using novel high-fidelity computational tools that are accurate and fast, and therefore can be used for parametric studies and uncertainty quantification. To sufficiently demonstrate the accuracy of these tools, advanced research reactors are needed. The authors argue the need for new computational paradigms such as the MRT (Multistage, Response- function Transport) methodology which has resulted in the development of the novel high-fidelity RAPID (Real-time Analysis for Particle-transport and In-situ Detection) code system. Such code systems have to be robust in modeling any complex system, and should be fast and accurate, henceforth their uncertainties can be quantified at reasonable costs. Again, advanced research reactors are needed for the validation of the fidelity and accuracy of new computational tools. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Computational simulation of a 100 kW dual circulating fluidized bed reactor processing coal by chemical looping with oxygen uncoupling

Chemical looping with oxygen uncoupling (CLOU) is a promising carbon capture technology that utilizes two interconnected fluidized-bed reactors to separate oxygen from air using a metal oxide, and then to use that oxygen to combust fuel in a nitrogen-free environment. CLOU is particularly well suited for solid fuels such as coal since released O 2 will readily combust with solid char. This study presents simulations of the University of Utah's dual fluidized bed chemical looping process development unit (PDU) operating as a CLOU system at 100 kW th , using CPFD Software's Barracuda-VR™ CFD modeling program. A full 3D model of the PDU was used including both reactors, as well as loop seals and cyclone separators. Hydrodynamic settings and the drag model are based on previous cold-flow validation experiments. Mechanisms describing CLOU reaction kinetics for cuprous oxide oxidation in the air reactor and cupric oxide reduction in the fuel reactor were adapted from previously completed experimental work. Coal char kinetics were based on the carbon-hydrogen-nitrogen-oxygen-sulfur, CHNOS, method. Submodels for devolatilization, gasification, and homogenous gas-phase reactions used relationships from the literature. Results predict volatile burnout almost immediately, though coal combustion continues into the upper portions of the fuel reactor and cyclone. Lastly, nitrogen gas leakage into the fuel reactor is shown to be mostly carryover from the air reactor

42 ENGINEERING↗
DOE OSTI2018

Heat Storage Options for Sodium, Salt and Helium Cooled Reactors to Enable Variable Electricity to the Grid and Heat to Industry with Base-Load Reactor Operations

Electricity markets are changing rapidly because of (1) the addition of wind and solar and (2) the goal of a low-carbon electricity grid. These changes result in times of high electricity prices and very low or negative electricity prices. California has seen its first month where more than 20% of the time (mid-day) the wholesale price of electricity was zero or negative. This creates large incentives for coupling heat storage to advanced reactors to enable variable electricity and industrial-heat output (to maximize revenue) while the reactor operates at base load (minimize cost). Recent studies have examined coupling various types of heat storage to Rankine and Brayton power cycles. However, there has been little examination of heat-storage options between (1) the reactor and (2) the power-conversion system or industrial customer. Heat-storage systems can be incorporated into sodium , helium-, and salt-cooled reactors. Salt-cooled reactors include the fluoride-salt-cooled high-temperature reactor (FHR) with its solid fuel and clean coolant and the molten salt reactor (MSR) with its fuel dissolved in the salt. For sodium and salt reactors, it is assumed that a heat-storage system would be in the secondary loop between the reactor and power cycle. For helium-cooled reactors, heat storage can be in the primary or secondary loop. This report is a first look at the rational and the heat storage options for deploying gigawatt-watt hour heat-storage systems with GenIV reactors. Economics and safety are the primary selection criteria. The leading heat-storage candidate for sodium-cooled systems (a low-pressure secondary system with small temperature drop across the reactor core) is sodium coolant stored in large steel tanks, with the coolant transferring heat in channels. The design minimizes sodium volume in the storage and, thus, the risks and costs associated with sodium. For helium systems (high-pressure with large temperature drop across the core), the leading heat storage options are (1) varying the temperature of the reactor core, (2) steel or alumina firebrick in a secondary pressure vessel and (3) nitrate or hot-rock/firebrick at atmospheric pressure. For salt systems (low pressure, high temperatures, and small temperature drop across the reactor core) the leading heat-storage systems are secondary salts. In each case, options are identified and questions to be addressed are identified. In some cases there is a strong coupling between the heat-storage technology and the power cycle. The leading sodium heat-storage technology may imply changes in the power cycle. High-temperature salt systems couple efficiency to Brayton power cycles that may create large incentives for the heat storage to remain within the power cycle rather than in any intermediate heat transfer loop.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Core design and safety analysis

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor II (EBR-II) for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor technologies. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Core design characteristics, fuel cycle performance, and safety analysis of the FHR DR preconcept have been evaluated. The FHR DR core design features a negative or negligible void coefficient throughout a reactor operating cycle. Both single-batch (cartridge) and multiple-batch fuel cycles can be demonstrated in the FHR DR. The single-batch cycle length of the FHR DR core is estimated at between 12-18 months, assuming the successful qualification of composite carbon (C/C) or silicon carbide (SiC/SiC) structural fuel block tie rod material. Fuel cycle performance of the FHR DR is similar to a high temperature gas-cooled reactor. Preliminary safety analysis of the FHR DR indicates that the reactor could be used to demonstrate the inherent safety characteristics of FHR designs.

Brown, Nicholas R.↗